Text

Rev. 1 to Holtec Report HI-2094289, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center.
	ML091940178
Person / Time
Site:	Indian Point
Issue date:	07/01/2009
From:	; Holtec
To:	; Office of Nuclear Reactor Regulation
References
	NL-09-076 HI-2094289, Rev. 1
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ENCLOSURE 2 TO NL-09-076 NON-PROPRIETARY VERSION OF THE PROPRIETARY DOCUMENT IN ENCLOSURE 1 Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

LICENSING REPORT ON THE INTER-UNIT TRANSFER of SPENT NUCLEAR FUEL at THE INDIAN POINT ENERGY CENTER By Holtec International Holtec Center 555 Lincoln Drive West Marlton, NJ 08053 (holtecintemational.com)

Holtec Project 1775 Holtec Report No. HI-2094289 Safety Category: Safety Significant Copyright Notice This document is a copyrighted initellectual property of 1-oltec International. 'All rights reserved. Ex~cerpting t o id , except for pblic domain citations iilided herein, by any person or entity except for the SNRC, a Htc User Group (HUG) member company, or a foreign regulatory authority with jurisdiction ove a UG':Merber's nula facility without written consent of 1{foltec International is unlawfiul,

HOLTEC INTERNATIONAL DOCUMENT NUMBER: 2094289 PROJECT NUMBER: 1775 DOCUMENT ISSUANCE AND REVISION STATUS DOCUMENT NAME: _Licensing Report for transfer of IP-3 fuel to IP-2 DOCUMENT CATEGORY: R] GENERIC Z PROJECT SPECIFIC REVISION No. 0 REVISION No. 1 REVISION No.

Author's Date VIR # Author's Date VIR #

Author's Date VIR #

No. cent Initials Approved Initials Approved Initials Approved

1. Chap 1 TSM 5/29/2009 434723 TSM 7/1/2009 836933

2. Chap 2 TSM 5/29/2009 59458 TSM 7/1/2009 685850

3. Chap 3 TSM 5/29/2009 765725 TSM 7/1/2009 407780

4. Chap 4 SPA 5/29/2009 341828 SPA 7/1/2009 465884

5. Chap 5 DMM 5/29/2009 541521 DMM 7/1/2009 90931

6. Chap 6 CWB 5/29/2009 140206 VRP 7/1/2009 533871

7. Chap 7 CS 5/29/2009 261166 CS 7/1/2009 60138

8. Chap 8 VG 5/29/2009 488015 VG 7/1/2009 159807

9. Chap 9 KAP 5/29/2009 850390 KAP 7/1/2009 773001

10. Chap 10 TSM 5/29/2009 873968 KAP 7/1/2009 421127

11. Chap 11 TSM 5/29/2009 206569 TSM 7/1/2009 86235

12. TOC TSM 5/29/2009 279864 VG 7/1/2009 868752 tt Chapter or section number.

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HOLTEC INTERNATIONAL DOCUMENT NUMBER: 2094289 PROJECT NUMBER: 1775 DOCUMENT CATEGORIZATION In accordance with the Holtec Quality Assurance Manual and associated Holtec Quality- Procedures (HQPs), this document is categorized as a:

ED Calculation Package 3 (Per HQP 3.2) [ Technical Report (Per HQP 3.2)(Such as a Licensing Report)

LI Design Criterion Document (Per HQP 3.4) D Design Specification (Per HQP 3.4)

El Other (Specify):

DOCUMENT FORMATTING The formatting of the contents of this document is in accordance with the instructions of HQP 3.2 or 3.4 except as noted below:

DECLARATION OF PROPRIETARY STATUS IN Nonproprietary El Holtec Proprietary EL Privileged Intellectual Property (PIP)

Documents labeled Privileged Intellectual Property contain extremely valuable intellectual/commercial property of Holtec International. They cannot be released to external organizations or entities without explicit approval of a company corporate officer. The recipient of Hloltec's proprietary or Top Secret document bears full and undivided responsibility to safeguard it against loss or duplication.

Notes:

1. This document has been subjected to review, verification and approval process set forth in the 1-loltec Quality Assurance Procedures Manual. Password controlled signatures of Holtec personnel who participated in the preparation, review, and QA validation of this document are saved in the N-drive of the company's network. The Validation Identifier Record (VIR) number is a random number that is generated by the computer after the specific revision of this document has undergone the required review and approval process, and the appropriate Holtec personnel have recorded their password-controlled electronic concurrence to the document.

2. A revision to this document will be ordered by the Project Manager and carried out if any of its contents is materially affected during evolution of this project. The detemfination as to the need for revision will be made by the Project Manager with input from others, as deemed necessary by him.

3. Revisions to this document may be made by adding supplements to the document and replacing the "Table of Contents", this page and the "Revision Log".

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LICENSING REPORT REVISION STATUS, LIST OF AFFECTED CHAPTERS AND REVISION

SUMMARY

License Report No.: H1-2094289 Revision Number: I Report

Title:

License Report for Fuel Transfer from Unit 3 SFP to Unit 2 SFP This Report is submitted to the USNRC in support of the LAR for fuel transfer from Indian Point Unit 3 to Indian Point Unit 2 spent fuel pool.

Chapter review and verification are controlled at the chapter level and changes are annotated at the Chapter level.

A summary description of change is provided below for each chapter section. Chapter sections not listed remain at the previous revision level. Minor editorial changes to this Report are not described.

Current Chapter Revision Summary Description of Change Noo.

TOC 1 This is a complete revision therefore no revision bars are shown 1 1 This is a complete revision therefore no revision bars are shown 2 1 This is a complete revision therefore no revision bars are shown 3 1 This is a complete revision therefore no revision bars are shown 4 1 This is a complete revision therefore no revision bars are shown 5 1 This is a complete revision therefore no revision bars are shown 6 1 This is a complete revision therefore no revision bars are shown Page S I of S2

7 1 This is a complete revision therefore no revision bars are shown 8 1 This is a complete revision therefore no revision bars are shown 9 1 This is a complete revision therefore no revision bars are shown 10 1 This is a complete revision therefore no revision bars are shown 11 1 This is a complete revision therefore no revision bars are shown Page S2 of S2

Table of Contents GLO SSA RY .................................................................................................................................... v 1.0 Introduction ............................................................................................................................ 1-1 1.1 Background and Overview .............................................................................................. 1-1 1.2 Sum m ary of Proposed Action .......................................................................................... 1-6 1.3 Description of Required Equipm ent and their Safety Function ....................................... 1-7 1.3.1 Shielded Transfer Canister .................................................................................... 1-7 1.3.2 H I-TRA C 100D Transfer Cask .............................................................................. 1-8 1.3.3 V ertical Cask Transporter ...................................................................................... 1-9 1.4 Inter-U nit Transfer Operations ...................................................................................... 1-12 1.5 Reference Draw ings ....................................................................................................... 1-32 1.6 Supplier's Qualification ................................................................................................. 1-33 2.0 Fuel Acceptance Criteria and Engineered M easures for Safety ............................................ 2-1 2.1 Fuel Acceptance Criteria .................................................................................................. 2-1 2.2 Safety and Protective M easures ....................................................................................... 2-2 2.2.1 Criticality Safety through Physical D esign ............................................................ 2-2 2.2.2 Criticality Safety through A ssured Boron Concentration ...................................... 2-2 2.2.3 Release Protection by M ultiple Barriers ................................................................ 2-2 2.2.4 Protection by a Favorable Thermal-Hydraulic Environment ................................. 2-3 2.2.5 Protection by the Selection of Low Dose Em itting Fuel ........................................ 2-3 2.2.6 Protection by Use of Proven Equipm ent ................................................................ 2-3 2.2.7 Protection by M aterial Selection ............................................................................ 2-3 2.2.8 Reliability through Increased Structural M argins .................................................. 2-4 2.2.9 Protection by Design to Prevent Inadvertent W ater Loss ...................................... 2-4 3.0 Principal Design Criteria, Applicable Loads, and Postulated Accidents ............................... 3-1 3.1 Governing Regulatory Requirem ents .............................................................................. 3-1 3.1.1 Criticality ................................................................................................................ 3-1 3.1.2 Shielding ................................................................................................................. 3-2 3.1.3 Therm al .................................................................................................................. 3-3 3.1.4 Structural ................................................................................................................ 3-3 3.2 Applicable (Design, Normal, and Postulated Accident) Loadings .................................. 3-6 3.2.1 D esign Basis Loads ................................................................................................. 3-6 3.2.2 N orm al Condition Loads ........................................................................................ 3-6 3.2.3 Accident Condition Loads ...................................................................................... 3-6 3.2.4 Load Cases ........................................................................................................... 3-11 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 i Rev. I

4.0 Criticality Evaluation ............................................................................................................ 4-1 4.1 Introduction ...................................................................................................................... 4-1 4.2 M ethodology .................................................. .............. ............................... 4-1 4.3 Acceptance Criteria ......................................................................................................... 4-2 4.4 A ssumptions ..................................................................................................................... 4-4 4.5 Input Data ......................................................................... ................................................ 4-5 4.5.1 D esign Basis Fuel Assem bly Specification ............................................................ 4-5 4.5.2 Core Operating Param eters ................................................................................... 4-5 4.5.3 A xial Burnup Distribution ...................................................................................... 4-5 4.5.4 Fuel Basket & STC Specifications ......................................................................... 4-5 4.6 Com puter Codes ............................................................................................................. 4-15 4.7 Analysis .......................................................................................................................... 4-15 4.7.1 Fuel Assem blies, Burnable Poison and H afnium Inserts ............ ......................... 4-16 4.7.2 Reactivity Effect of Axial Burnup D istribution ................................................... 4-16 4.7.3 Isotopic Com positions .......................................................................................... 4-16 4.7.4 Uncertainty in D epletion Calculations ................................................................. 4-16 4.7.5 Eccentric Fuel Assembly Positioning.... ........................ 4-16 4.7.6 U ncertainties from Manufacturing Tolerances ..................................................... 4-17 4.7.7 Temperature & W ater D ensity Effects ................................................................ 4-18 4.7.8 Calculation of Maximum Keff for Normal Conditions ........................................ 4-19 4.7.9 Abnormal & Accident Conditions .............................. 4-19 4.7.10 A dditional Sensitivity Studies .............................................................................. 4-20 4.8 Acceptability of Storing IP3 fuel in IP2 pool ................................................................ 4-36 Appendix 4.A Appendix 4.B 5.0 Therm al-Hydraulic Evaluation .............................................................................................. 5-1 5.0 Overview .......................................................................................................................... 5-1 5.1 Therm al D esign ............................................................................................................... 5-4 5.1.1 O ver Pressure Protection ....................................................................................... 5-5 5.2 Therm al Properties of Materials ...................................................................................... 5-7 5.3 Thermal Evaluation of Fuel Transfer Operation ......................... 5-16 5.3.1 D escription of the 3-D Therm al M odel ................................................................ 5-16 5.3.2 M axim um Temperatures ................................................. I..I................................. 5-18 5.3.3 Evaluation of STC w ithout the HI-TRA C ............................................................ 5-19 5.4 Hypothetical Accident Evaluations ................................................................................ 5-24 5.4.1 Jacket W ater Loss Accident ................................................................................. 5-24 5.4.2 Fire Accident .............. .............................. 5-24

.6.0 Structural Evaluation of N orm al & A ccident Condition Loadings ....................................... 6-1 6.0 Overview .......................................................................................................................... 6-1 6.1 Structural D esign ............................................................................................................. 6-2 6.1.1 D iscussion ............................................................................................................... 6-2 6.1.2 D esign Criteria and Applicable Loads ................................................................... 6-2 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 ii Rev. 1

6.2 Structural Analyses ............................................... ;......................................................... 6-9 6.2.1 Load Case 1: D esign Pressure ................................................................................ 6-9 6.2.2 Load Case 2: Normal Operating Pressure plus Temperature ............................... 6-11 6.2.3 Load Case 3: N orm al H andling ............................................................................ 6-12 6.2.4 Load Case 4: Fuel A ssembly Drop A ccident ........................................................ 6-16 6.2.5 Load Case 5: H I-TRA C V ertical Drop A ccident ................................................. 6-17 6.2.6 Load Cases 6 and 7: Seismic Stability of Loaded VCT and Loaded HI-TRAC... 6-19 6.2.7 Load Case 8: Seism ic Stability of STC in the Fuel Pool ...................................... 6-20 7.0 Shielding D esign & ALA RA Considerations ........................................................................ 7-1 7.0 Introduction ...................................................................................................................... 7-1 7.1 Shielding Design .............................................................................................................. 7-3 7.1.1 D esign Features ...................................................................................................... 7-3 7.1.2 A cceptance Criteria ................................................................................................ 7-3 7.2 Source Specification ....................................................................................................... 7-5 7.2.1 Source Term Selection ........................................................................................... 7-5 7.2.2 Principal Sources of Radiation ............................................................................... 7-5 7.3 Shielding Model ............................................................................................................. 7-10 7.3.1 Configuration of Shielding and Source ................................................................ 7-10 7.3.2 M aterial Properties ............................................................................................... 7-11 7.4 Shielding and ALA RA Evaluation ................................................................................ 7-16 7.4.1 M ethods ................................................................................................................ 7-16 7.4.2 Flux-to-D ose-Rate Conversions ........................................................................... 7-18 7.4.3 External Radiation Levels - STC ......................................................................... 7-18 7.4.4 External Radiation Levels - HI-TRAC IOOD Transfer Cask ............................... 7-18 7.4.5 D ose Contribution to Site Boundary .................................................................... 7-18 7.4.6 Effluent Dose Evaluation ..................................................................................... 7-19 7.4.7 O ccupational Exposures for ALA RA Considerations ......................................... 7-20 7.4.8 Summary and Conclusions ................................................................................... 7-20 8.0 Materials Evaluation, Acceptance Tests and Maintenance Program ..................................... 8-1 8.1 Introduction ...................................................................................................................... 8-1 8.2 M aterials U sed ................................................................................................................. 8-1 8.3 Degradation M echanism s ................................................................................................. 8-6 8.4 Acceptance Tests ............................ ......... ...... .......... 8-10 8.4.1 V isual Inspections and M easurem ents ................................................................. 8-10 8.4.2 W eld Exam inations .............................................................................................. 8-10 8.4.3 Structural and Pressure Tests ............................................................................... 8-11 8.4.4 STC Leakage Test ................................................................................................ 8-11 8.4.5 Com ponent and M aterial Tests ............................................................................ 8-11 8.5 M aintenance Program .................................................................................................... 8-12 8.5.1 Structural and Pressure Tests .............................................................................. 8-12 8.5.2 STC Leakage Test ................................................................................................ 8-12 8.5.3 Component and M aterial Tests ............................................................................ 8-12 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 iii Rev. I

9.0 Economic and Environmental Considerations ....................................................................... 9-1 9.1 Environmental Considerations ......................................................................................... 9-1 9.1.1 Environmental Impacts .......................................................................................... 9-2 9.1.2 Occupational Radiation Exposure .......................................................................... 9-3 9.1.3 Public Radiation Exposure ..................................................................................... 9-3 9.2 Economic Considerations ................................................................................................ 9-4 9.2.1 Spent Fuel Storage Options Evaluated ............... ............. 9-4 9.2.2 Options to Implement Dry Cask Storage at IP-3 .................................................... 9-5 10.0 Operating Procedures ..................................................................................................... 10-1 10.0 Introduction .................................................................................................................. 10-1 10.1 STC Preparation and Setup .......................................................................................... 10-3 10.1.1 STC Inspections and Checkout ........................................................................... 10-3 10.1.2 HI-TRAC Inspections and Checkout .................................................................. 10-3 10.1.3 Preparation and Setup for Use ............................................................................ 10-3 10.2 STC Fuel Loading ........................................................... ............................................. 10-5 10.2.1 Placement of STC in the SFP ............................................................................. 10-5 10.2.2 Fuel Handling and Loading into STC ................................................................. 10-6 10.2.3 Removal of STC from SFP and placement in HI-TRAC ................................... 10-8 10.3 HI-TRAC/STC Transfer ........................................................................................... 10-10 10.3.1 Haul Path Inspection and Controls ................................................................... 10-11 10.3.2 M ovement of loaded HI-TRAC on LPT ........................................................... 10-11 10.3.3 M ovement of loaded HI-TRAC with VCT ....................................................... 10-12 10.4 STC Fuel Unloading .......................................... 10-13 10.4.1 Placement of loaded STC in SFP ...................................................................... 10-13 10.4.2 Unloading of STC ............................................................................................ 10-14 10.4.3 Removal of STC from SFP and placement in HI-TRAC ............... 10-14 10.5 M aintenance and Off- Normal Events ....................................................................... 10-16 10.5.1 Crane Operational Event .................................. :................................................ 10-16 10.5.2 STC W ater Inventory Control ........................................................................... 10-16 10.5.3 VCT Breakdown .............................................................................................. 10-16 10.5.4 Vertical Cask Drop Recovery Plan ................................................................... 10-16 10.5.5 Potential Damaged Fuel Assembly ................................................................... 10-17

,11.0 References..................................................................................................................... 11-1 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 iv Rev. 1

GLOSSARY AFR is an acronym for Away From Reactor ALARA is an acronym for As Low As Reasonably Achievable.

Cask is a generic term used to describe a device that is engineered to hold high level waste, including spent nuclear fuel, in a safe configuration.

C.G. is an acronym for Center of Gravity.

Commercial Spent Fuel (CSF) refers to nuclear fuel used to produce energy in a commercial nuclear power plant.

Cooling Time (or post-irradiation decay time, PCDT) for a spent fuel assembly is the time between reactor shutdown and the time the spent fuel assembly is placed in a cask system.

Cooling Time is also referred to as the "age" of the CSF.

Critical Characteristic means a feature of a component or assembly that is necessary for the component or assembly to render its intended function. Critical characteristics of a material are those attributes that have been identified, in the associated material specification, as necessary to render the material's intended function.

DBE means Design Basis Earthquake.

DCSS is an acronym for Dry Cask Storage System.

Design Life is the minimum duration for which the component is engineered to perform its intended function.

Design Specification is a document prepared in accordance with the quality assurance requirements of 10CFR72 Subpart G to provide a complete set of design criteria and functional requirements for a system, structure, or component, designated as Important to Safety, intended to be used in the operation, implementation, or decommissioning of the cask system.

Equivalent (or Equal) Material is a material whose critical characteristics(see definition above) meet or exceed those specified for the designated material.

Fracture Toughness is a material property that is a measure of the ability of a material to limit crack propagation under a suddenly applied load.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 V Rev. 1

FSAR is an acronym for Final Safety Analysis Report under the regulations of 10CFR72 or 10CFR50. When not specifically designated, the FSAR referred to in this report is the HI-STORM 100 FSAR.

Fuel Basket means a honeycombed cavity structure with square openings that can accept a fuel assembly of the type for which it is designed.

High Burnup Fuel (HBF) is a commercial spent fuel assembly with an average burnup greater than 45,000 MWD/MTU.

HI-TRAC transfer cask or HI-TRAC means the transfer cask which is used to hold the STC during the inter unit transfer of fuel assemblies from Unit 3 spent fuel pool to Unit 2 spent fuel pool.

Important to Safety (ITS) means a function or condition required to store spent nuclear fuel safely; to prevent damage to spent nuclear fuel during handling and storage, and to provide reasonable assurance that spent nuclear fuel can be received, handled, packaged, stored, transferred and retrieved without undue risk to the health and safety of the public.

Incore Grid Spacers are fuel assembly grid spacers located within the active fuel region (i.e.,

not including top and bottom spacers).

Intact Fuel Assembly is defined as a fuel assembly without known or suspected cladding defects greater than pinhole leaks and hairline cracks, and which can be handled by normal means. Fuel assemblies without fuel rods in fuel rod locations shall not be classified as Intact Fuel Assemblies unless dummy fuel rods are used to displace an amount of water greater than or equal to that displaced by the original fuel rod(s).

Inter-unit transfer means transfer of SNF from IP3 fuel pool to IP2 fuel pool IP-2 means Indian Point Unit 2 IP-3 means Indian Point Unit 3 LLNL is the acronym for Lawrence Livermore National Lab License Life means the duration for which the system is authorized by virtue of its certification by the U.S. NRC.

Light Water Reactor (LWR): Reactors that utilize enriched uranium and/or the so-called MOX fuel for power generation.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 vi Rev. I

Low Profile Transporter (LPT) is a device for moving a loaded HI-TRAC into Unit 2 building.

Lowest Service Temperature (LST) is the minimum metal temperature of a part for the specified service condition.

Maximum Reactivity means the highest possible k-effective including bias, uncertainties, and calculational statistics evaluated for the worst-case combination of fuel basket manufacturing tolerance.

Metamic is a trade name for an aluminum/boron carbide composite neutron absorber material qualified for use in the MPC fuel baskets and spent fuel racks. Metamic is used in the STC basket.

Minimum Enrichment is the minimum assembly average enrichment. Natural uranium blankets are not considered in determining minimum enrichment.

Moderate Burnup Fuel (MBF) is a commercial spent fuel assembly with an average burnup less than or equal to 45,000 MWD/MTU.

Multi-Purpose Canister (MPC) means the sealed canister consisting of a honeycombed fuel basket for spent nuclear fuel storage, contained in a cylindrical canister shell (the MPC Enclosure Vessel).

NDT is an acronym for Nil Ductility Transition, which is defined as the temperature at which the fracture stress in a material with a small flaw is equal to the yield stress in the same material if it had no flaws.

Neutron Shielding means a material used to thermalize and capture neutrons emanating from the radioactive spent nuclear fuel.

Neutron Sources means specially designed inserts for fuel assemblies that produce neutrons for startup of the reactor.

Non-Fuel Hardware (NFH) is defined as Burnable Poison Rod Assemblies (BPRAs), Thimble Plug Devices (TPDs), Control Rod Assemblies (CRAs), Axial Power ShapingRods (APSRs),

Wet Annular Burnable Absorbers (WABAs), Rod Cluster Control Assemblies (RCCAs),Control Element Assemblies (CEAs), Neutron Source Assemblies (NSAs), water displacement guide tube plugs, orifice rod assemblies, vibration suppressor inserts, instrument tube tie-rod and components of these devices such as individual rods.

Not-Important-to-Safety (NITS) is the term used where a function or condition is not deemed as Important-to-Safety. See the definition for Important-to-Safety.

ORNL is the acronym for Oak Ridge National Laboratory.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 vii Rev. I

Planar-Average Initial Enrichment is the average of the distributed fuel rod initial enrichments within a given axial plane of the assembly lattice.

Post-Core Decay Time (PCDT) is synonymous with cooling time.

PWR is an acronym for Pressurized Water Reactor.

Reactivity is used synonymously with effective neutron multiplication factor or k-effective.

Regionalized Fuel Loading is a term used to describe an optional fuel loading strategy used in lieu of uniform fuel loading. Regionalized fuel loading allows high heat emitting fuel assemblies to be stored in fuel storage locations in the center of the fuel basket provided lower heat emitting fuel assemblies are stored in the peripheral fuel storage locations.

Service Life means the duration for which the component is reasonably expected to perform its intended function. Service Life may be much longer than the Design Life because of the conservatism inherent in the codes, standards, and procedures used to design, fabricate, operate, and maintain the component.

Short-term Operations means those normal operational evolutions necessary to support fuel loading or fuel unloading operations.

Single Failure Proof means that the handling system is designed so that a single failure will not result in the loss of the capability of the system to safely retain the load. A Single Failure Proof means that the handling system is designed so that all directly loaded tension and compression members are engineered to satisfy the enhanced safety criteria of Paragraphs 5.1.6(l)(a) and (b) of NUREG-0612.

SNF is an acronym for Spent Nuclear Fuel (also referred to as CSF).

SSC is an acronym for Structures, Systems and Components.

STP is Standard Temperature (298°K) and Pressure (1 atm) conditions.

Surface Contaminated Object (SCO) means a solid object that is not itself classed as radioactive material, but which has radioactive material distributed on any of its surfaces. See 10CFR71.4 for surface activity limits and additional requirements.

Thermosiphon is the term used to describe the buoyancy-driven natural convection circulation of Coolant to reduce the temperature of the spent nuclear fuel.

Shielded Transfer Canister (STC) means a thick walled cylindrical container that is compatible with the HI-TRAC transfer cask and serves as the enclosure for wet transfer of the IP3 fuel to IP2 pool.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 viii Rev. 1

Vertical Cask Transporter (VCT) is used for vertical handling and on-site moving of the loaded or empty HI-TRAC transfer cask.

ZPA is an acronym for Zero Period Acceleration.

ZR means any zirconium-based fuel cladding material authorized for use in a commercial nuclear power plant reactor. Any reference to Zircaloy fuel cladding applies to any zirconium-based fuel cladding material.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 ix Rev. I

CHAPTER 1: INTRODUCTION 1.1 Background and Overview Indian Point Unit 2 (IP-2) and Indian Point Unit 3 (IP-3) are Westinghouse Pressurized Water Reactors (PWR) co-located in Buchanan, NY, approximately 40 miles north of New York City.

The two plants IP-2 and IP-3 were designed by the same architect/engineer and were built by the same construction company during the same period. Both plants operate using the same PWR fuel assembly type. Together, IP-2 and IP-3 comprise the Indian Point Energy Center (IPEC),

owned and operated by Entergy. Both units' spent fuel pools (SFP) were re-racked in the late 1980s to maximize their in-pool storage capacities.

IP-2, which began operation two years prior to IP-3, began transferring its Spent Nuclear Fuel (SNF) from the SFP into dry storage employing the HI-STORM 100 System (USNRC Docket 72-1014) at an on-site Independent Spent Fuel Storage Installation (ISFSI) in 2007. As of April 2009, 96 spent fuel assemblies have been placed in dry storage in three (3) HI-STORM 100 Systems. Additional transfers from wet storage into dry storage are planned to be carried out at approximately one year intervals. It should also be noted that in August 2008, 160 fuel assemblies from the Indian Point Unit 1 SFP were also placed into dry storage at this ISFSI using five (5) modified HI-STORM 100 Systems (see Certificate of Compliance 1014, Amendment 4).

The IP-2 Fuel Storage Building (FSB) required a significant structural upgrade to make it suitable for placing the fuel into dry storage. Prior to the start of the project the building crane was rated at 40 tons and was classified as non-single failure-proof. A facility must have a crane rated to at least 100 tons to load the multi-purpose canisters (MPCs) which contain 32 PWR fuel assemblies, and all lifts are required to be either single failure proof or supplemented with redundant drop protection features designed to protect against heavy load drops. Upgrading the IP-2's crane capacity to 100 tons and to single-failure proof required a whole new gantry, trolley and control system, along with a whole new sub-grade support system. The resulting structural modification effort in the IP-2 FSB was both long and costly and involved hard-rock excavations immediately adjacent to safety-significant structures and equipment. The cask loading infrastructure at IP-2 is now operationally sound and proven through repeated use.

The need to begin defueling the IP-3 SFP is now imminent, i.e. a minimum of 96 fuel assemblies must be removed from the IP-3 SFP by the end of 2010 to restore the full core reserve capability prior to the spring 2011 refueling outage and ability to receive new fuel.

As the considerations of credible options for achieving the reduction of IP-3's inventory in Chapter 9 indicate, the wet transfer of fuel from the IP-3 SFP to the IP-2 SFP is the safest, most consistent with ALARA, most environmentally benign, and most economical option. It has been determined that a wet transfer from the IP-3 SFP to the IP-2 SFP, with a maximum of 12 fuel assemblies per transfer, can be carried out in full compliance with all safety predicates of 10 CFR 50 without undertaking a major structural modification of the IP-3 FSB in the manner of IP-

2. The existing 40 ton crane which is currently not single failure proof will be replaced with a 40 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-1 Rev. I

ton single failure proof crane, and the loading on the building structure will remain within the originally engineered limits.

A Shielded Transfer Canister (STC) has been designed to hold the fuel in a wet environment while being transferred between the IP-3 SFP and the IP-2 SFP. The STC will be moved a distance of about 300 yards between the IP-3 and IP-2 FHBs using two major pieces of equipment that have been successfully used to move loaded Holtec MPCs containing 32 spent fuel assemblies for the IP-2 dry storage program. These are:

i. the HI-TRAC 1OOD Transfer Cask (HI-TRAC), and ii. the on-site Vertical Cask Transporter (VCT).

Therefore, the STC and any supporting ancillary equipment are the only new pieces of capital equipment required for the inter-unit fuel transfer.

The Shielded Transfer Canister is placed in the HI-TRAC for movement from IP-3 to IP-2 by the VCT at the speed of approximately 1/2 mile per hour (typical of on-site transporters). The unique features of this transfer are:

i. The STC features a Code-compliant, bolted closure lid which is sealed using an elastomeric seal. This seal will be tested to ensure there is no leakage of water through the seal. The fuel in the STC is in the same thermal-hydraulic environment as the pool (i.e., no risk of thermal shock to the fuel).

ii. The top of the HI-TRAC has a solid, bolted lid which is sealed using an elastomeric seal. It is designed to meet the stress limits of the ASME Code,Section III, Subsection NF. The bottom lid of HI-TRAC is also equipped with an elastomeric seal.

This packaging arrangement has the explicit consequence of establishing a high-integrity barrier against the leakage of contaminated water to the open air. In fact as the information assembled in Table 1.1.1 indicates, the loaded STC placed in the HI-TRAC and moved by the VCT provides a level of protection that is comparable to that engineered in the plant's wet storage system.

Prior to the compilation of this report, it was necessary to identify the regulatory regimen under which NRC's approval of these planned activities should be requested. The licensing organizations of Holtec International and Entergy concluded that the proposed activity most appropriately belongs to the amendment process under 10CFR 50. The alternative of using Part 72 is inappropriate because Part 72 is applicable only to a program that entails dry fuel storage at an ISFSI. The alternative of securing a Part 71 transportation certificate was also considered and rejected because such a license applies to handling and preparing a package intended for the transportation of radioactive material, outside the confines of a licensee's facility or authorized place of use. On the other hand, an in-depth review of the work effort in the proposed inter-unit transfer indicated the requirements of 10CFR 50 apply. The principal essential operational HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-2 Rev. I

characteristics of the inter-unit transfer project that make it necessary to seek an amendment under Part 50 are:

i. The fuel is always maintained in a wet environment.

ii. The criticality safety of fuel in a wet environment is fully addressed in Part 50.

iii. The heavy load handling evolutions will be confined to the Part 50 structures of the two FSBs (both within the purview of Part 50).

iv. All credible fuel handling accidents and their consequence to the plants' charcoal filter's effectiveness and HVAC system must be treated in accordance with NRC publications such as the Generic Letter 78-11 OT Position paper [F.A].

v. No activity will occur outside the site's protected area. As required under 10CFR50, the security provisions of Part 73 and SNM control per Part 74 will apply without limitation.

vi. The transfer of the spent nuclear fuel assemblies between the two fuel pools is permitted under 10CFR 70.42, which is compatible with the overarching requirements of 10CFR 50.

However, in addition to mandating 10CFR 50 regulations, supplemental requirements from other regulations such as Part 72 are invoked if a specific requirement is not explicitly provided in the Part 50 documents and is determined by Holtec International and Entergy to be desirable for reinforcing the design and acceptance criteria germane to the safety features of the planned work effort.

This Licensing Report provides a summary of all analyses and evaluations performed to establish that the inter-unit transfer process meets all criteria discussed in Chapter 3.

The object of this Licensing Report is to provide the substantiating information to NRC NRR in support of two amendments as follows:

i. An amendment to IP-3 Technical Specifications to load spent fuel into the STC for transfer to IP-2 and; ii. An amendment to IP-2 Operating License and Technical Specifications to receive and unload the fuel from the STC.

A series of chapters in this Licensing Report provide the necessary technical information in support of these amendments.

The following convention is used in the organization of chapters:

i. A chapter is identified by a whole numeral, e.g. Chapter 3.

ii. A section is identified by two numerals separated by one decimal, e.g., Section 3.1 is a section in Chapter 3.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-3 Rev. 1

iii. A subsection is identified by three numerals separated by two decimals, e.g., Subsection 3.2.1 is a subsection in Section 3.2.

iv. A paragraph is identified by four numerals separated by three decimals, e.g., Paragraph 3.2.1.1 is a paragraph in Subsection 3.2.1.

v. All figures, tables and references cited are identified by three numerals separated by two decimals, m.n.i, where "m" is the chapter number, "n" is the section number, and "i" is the sequential number. For example, Figure 1.2.3 is the third figure in Section 1.2 of Chapter 1.

Revisions to this document are made at the chapter level. Tables and figures associated with a section are placed after the text narrative. Complete chapters are replaced if any material in the chapter is changed. The specific changes are appropriately annotated with revision bars in the right margin. Drawing packages are controlled separately within the Holtec QA program and have individual revision numbers. If a drawing is revised in support of the current revision, that drawing is included in Section 1.5 at its latest revision level.

Chapter 11 contains the generic industry and Holtec produced references which may have been consulted in the preparation of this document. Where specifically cited, the identifier is listed in the text or table as "[A.A]". Active Holtec Calculation Packages which are the repository of all relevant licensing and design basis calculations are annotated as "latest revision".

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-4 Rev. I

Table 1.1.1 ENGINEERED SAFETY FEATURES INCLUDING BARRIERS BETWEEN THE FUEL AND THE AMBIENT ENVIRONMENT Engineered Feature During Inter-Unit During Storage in Transfer in STC and the Spent Fuel Pool HI-TRAC 100D Transfer Cask

1. Is the fuel surrounded by Borated water Yes Yes to mitigate reactivity?

2. Is the fuel protected from external Yes, by the HI-TRAC Yes, by the Fuel environmental loadings such as tornado Transfer Cask Storage Building missiles?

3. Is the fuel maintained in a cooled state? Yes, by Borated water Yes, by Borated water at equilibrium below and Fuel Pool Cooling saturation temperature and Cleanup System

4. Is there a risk of an uncontrolled No, the cask crane N/A lowering of the load in the Fuel Storage will be single-failure-Building? proof.

5. Is there a credible accident scenario for No, evaluated in this Yes, evaluated in the boiling of water? Licensing Report plant's UFSAR HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-5 Rev. I

1.2 Summary of Proposed Action Entergy / Indian Point Energy Center will transfer fuel assemblies from the IP-3 spent fuel pool (SFP) to the IP-2 SFP. The amount of spent nuclear fuel (SNF) in the SFP at IP-3 has reached levels prohibiting a full core off-load.

The crane limitation at IP-3 does not allow the use of a dry storage system currently approved under 10CFR 72 since the IP-3 crane is currently rated at 40 tons and the minimum requirement for dry storage is 100 tons (for the HI-STORM 100 System).

Due to structural limitations of the Fuel Storage Building (FSB) the crane cannot be upgraded to support this weight without a massive building structural modification which would include significant demolition immediately adjacent to the spent fuel pool structure.

The system described in the next section has been developed which uses equipment from Entergy's dry storage program approved for use under 10CFR 72, along with a newly designed equipment, a Shielded Transfer Canister (STC) and ancillaries, to perform a wet transfer of the fuel assemblies between the spent fuel pools.

The transferred IP-3 SNF will then be loaded into the IP-2 SFP and at a later date placed into dry storage. Since the IP-2 crane is a single failure proof crane rated at 100 tons it will not require further upgrades to perform this fuel transfer. The IP-3 crane will be upgraded to be single failure proof however the current crane capacity of 40 tons will not be increased. The new equipment is designed to allow for transferring a maximum number of assemblies while still maintaining the weight below the current IP-3 crane capacity.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-6 Rev. I

1.3 Description of Required Equipment and their Safety Function 1.3.1 Shielded Transfer Canister The Shielded Transfer Canister (STC) is a thick-walled cylindrical vessel with a welded base plate, and a bolted top lid. The internal cavity space of the STC is equipped with fuel basket to create twelve storage cells for transferring spent nuclear fuel (SNF) assemblies. The thicknesses of the STC shell, base plate, and top lid are substantially in excess of that required to meet their pressure retention function. As shown in Chapter 6, the pressure boundary of the STC meets the stress limits of ASME Boiler & Pressure Vessel Code Section III Class 3, Subsection ND with large margins and will not be a code stamped vessel. The applicable design pressure and temperature for the STC are listed in Table 3.2.1. According to the ASME B&PV Code Section ND-7000, pressure vessels are required to have overpressure protection; however no overpressure protection is provided in the STC. The function of the STC is to retain the radioactive contents under normal, off-normal, and accident conditions. The STC is designed to withstand a maximum internal pressure considering maximum accident temperatures and therefore does not require the pressure relief valve.

The STC has two lift points which will attach to the overhead cranes at IP-3 and IP-2 through the STC lid and a lifting device. The STC lifting points are designed in accordance with NUREG-0612 [C.A] for critical loads. The lid attaches using threaded studs and nuts.

The design of the STC is presented in the Licensing Drawings in Section 1.5. A cut-away view is shown in Figure 1.3.1. Other essential design characteristics of the STC are:

i. The basket is an open-ended honeycomb configuration of austenitic stainless steel plates with panels of Metamic neutron absorber for criticality control affixed to the plates under thin stainless steel sheathing. The edges of the plates are welded to each other.

ii. The honeycomb basket stands upright in the STC cavity. The peripheral space between the basket and the Shielded Transfer Canister inner wall is equipped with appropriate lateral spacers to ensure a consistent spacing between the basket and the inner wall of the STC. The area is not large enough to permit an inadvertent loading of a fuel assembly into the peripheral space.

iii. The special lifting device used to lift the STC meets the guidance of NUREG-0612, Section 5.1.6(1) and ANSI N14.6-1993 [B.S] for critical loads. The interfacing lift points are designed to meet the requirements of NUREG-0612, Section 5.1.6(3).

iv. The lifting attachments are engineered for remote engagement and disengagement to minimize time and dose.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-7 Rev. I

v. The arrangement of stainless steel and Metamic plates in the STC fuel basket is based on Holtec's multi-purpose canister, MPC-32, certified in Docket No. 72-1014, and used in numerous PWR dry storage applications in the U.S.(including IP-2).

The governing quality assurance requirements for design and fabrication of the STC are stated in 10CFR 50 Appendix B. Holtec's Nuclear Quality Assurance program (USNRC Docket No. 71-0784) complies with this regulation and is designed to provide a flexible but highly controlled system for the design, analysis, licensing and fabrication of customized components in accordance with the applicable codes, specifications, and regulatory requirements.

1.3.2 HI-TRAC IOOD Transfer Cask The STC will be placed in the HI-TRAC 100D Transfer Cask (HI-TRAC) which is an existing piece of equipment already used at IP-2 to transfer SNF from the SFP into dry storage. The HI-TRAC has a steel, lead, steel layered cylindrical shell for gamma radiation shielding with an outer annulus which can be filled with water for neutron shielding. The structural integrity is provided by the carbon steel. The lifting trunnions on the HI-TRAC are designed to meet the design safety factors of ANSI N14.6 for critical loads. The HI-TRAC is part of the HI-STORM 100 Dry Cask Storage System licensed under NRC Docket 72-1014 and is described fully in its Final Safety Analysis Report [K.A].

A solid top lid, designed for the HI-TRAC 100)D, has an elastomeric seal to retain the water present in the STC/HI-TRAC annulus space. The lid will be attached with multiple bolts to provide the necessary bolt pull to maintain joint integrity. The HI-TRAC is designed to meet the stress limits of the ASME Code,Section III, Subsection ND, Class 3. According to the ASME B&PV Code Section ND-7000, pressure vessels are required to have overpressure protection; however no overpressure protection is provided in the HI-TRAC. The function of the HI-TRAC is to retain its contents under normal, off-normal, and accident conditions. The HI-TRAC is designed to withstand a maximum internal pressure considering maximum accident temperatures and therefore does not require the pressure relief valve. The HI-TRAC is designed to:

i. Provide maximum shielding to the plant personnel engaged in conducting "short-term operations" pertaining to inter-unit transfer.

ii. Provide protection to the STC and the SNF against extreme environmental phenomena loads, such as tornado missiles, during short-term operations.

iii. Serve as the container equipped with the appropriate lifting devices in full design compliance with NUREG-0612, Section 5.1.6.(3) and ANSI N14.6 to lift, move, and handle the STC, as required, to perform the short-term operations.

The above performance demands on the HI-TRAC are met by its design configuration as summarized below and presented in the licensing drawings in Section 1.5 of the HI-STORM 100 FSAR, Holtec Report HI-2002444 and in Section 1.5 of this report for the lid design.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-8 Rev. I

The HI-TRAC, as described in NRC Docket No. 72-1014, is principally made of carbon steel and lead. The cask consists of two major parts, namely (a) a multi-shell Cylindrical cask body, and (b) a multi-plate bottom lid. The cylindrical cask body is made of three concentric shells joined to a solid annular forging at their two extremities by circumferentially continuous welds.

The innermost and the middle shell are fixed in place by longitudinal connector ribs which serve as radial connectors between the two shells. The radial connectors provide a continuous path for radial heat transfer and render the dual shell configuration into a stiff beam under flexural loadings. The space between the two shells is occupied by lead, which provides the bulk of the cask's radiation shielding capability and accounts for a major portion of its weight.

Between the middle shell and the outermost shell is the outer annulus space that is referred to as the water jacket. This space is filled with uncontaminated water and provides most of the neutron shielding capability to the cask.

The bottom, of the Transfer Cask has a thick lid that makes the cask a watertight container using an elastomeric seal against the machined face of the forging. A set of bolts that tap into the machined holes in the lid provide the required physical strength to meet the structural requirements of ANSI 14.6 and to provide the necessary bolt pull to maintain joint integrity.

A cut-away view of the HI-TRAC IOOD is shown in Figure 1.3.2.

1.3.3 Vertical Cask Transporter The HI-TRAC will be lifted and moved using the IP-2 Vertical Cask Transporter (VCT). This is an existing piece of equipment which has been used at the site during the dry cask storage campaign licensed under 10 CFR 72. An existing low profile transporter (LPT) which has also been used during the above mentioned IP-2 dry storage campaign will be used to move the HI-TRAC into the IP-2 building. Air pads will be used at the IP-3 building to move the HI-TRAC out of the building before being placed on the VCT. The Vertical Cask Transporter is a high-capacity, tracked vehicle designed specifically for the lifting and handling of spent fuel storage components. The VCT lifts the HI-TRAC via special lifting devices designed, constructed and tested in accordance with ANSI N 14.6. The HI-TRAC is lifted using hydraulic lifting towers which have features to prevent a load drop even under complete hydraulic line failure. In addition, special locking pins secure the load during movement providing redundant drop protection. A hydraulically-tightened safety strap secures the cask in the VCT and prevents rocking or swaying of the cask under movement. Finally, the VCT is equipped with speed governing features, which limits the travel speed to approximately 0.5 mph and prevents coasting on a loss of power condition, and a braking system with emergency stop which overrides all other controls and brings the Vertical Cask Transporter to a stop.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-9 Rev. I

Figure 1.3.1: Cut-Away View of the Shielded Transfer Canister HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-10 Rev. i

Figure 1.3.2: Cut-Away.View of the H-TRAC 1OOD Transfer Cask HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-11 Rev. 1

1.4 Inter-Unit Transfer Operations The below operational steps are pictorially illustrated in Figures 1.4.1 through 1.4.17.

The STC is a thick-walled vessel with a removable top lid capable of transferring up to twelve IP-3 spent fuel assemblies. The STC is used in conjunction with the HI-TRAC transfer cask currently licensed under 10CFR Part 72 in Docket No. 72-1014 and used at IP-2. During STC closure and transfer, the STC shielding is supplemented with the HI-TRAC shielding (steel, lead and water) and the water contained in the annulus space located between the STC and the HI-TRAC. Since the outer diameter of the STC is smaller than the MPC used at IPEC, the STC is equipped with radial spacer ribs at the top and bottom. These ribs keep the STC centered inside HI-TRAC cavity and provide a soft edge lead-in during cask loading operations to help prevent damage to the inside coatings of HI-TRAC. The STC includes a removable bolted lid with penetrations for water filling and draining purposes. These penetrations are equipped with isolation valves which will remain attached to the STC lid. The lid also features threaded lid lifting points. The STC top end features cask lifting points which will provide a means to attach it to the overhead cranes. The spacer ribs form an annular region inside HI-TRAC which remains mostly full of water during loading and transfer operations. An air space is purposely left in the HI-TRAC above the height of the STC lid for two purposes; to allow the STC lid operations to occur unhindered by water, and to provide an expansion zone for the water inside the HI-TRAC cavity.

The STC is moved between IP-2 and IP-3 vertically in the HI-TRAC. Neither the HI-TRAC nor STC will be handled in the horizontal orientation when loaded. In addition to the internal STC cavity water and the water in the annulus space between the STC and HI-TRAC's inner shell, HI-TRAC's water jacket is also filled with water. These three discrete zones of water provide shielding and aid in heat transfer.

At the start of operations, the HI-TRAC top lid is removed and the empty STC is placed inside the HI-TRAC. The STC ribs center the STC inside of the HI-TRAC. The HI-TRAC's top lid is installed on the HI-TRAC to prevent any spilling of the water during the outside transfer process.

Movement of HI-TRAC (containing the STC) is performed using the Vertical Cask Transporter (VCT), the IP-2 Low Profile Transporter (LPT) or air-pads as described below in the synopsis of the fuel transfer operations.

The VCT moves the HI-TRAC containing the empty STC outside the IP-3 FSB truck bay door.

The HI-TRAC is lowered onto air-pads and the VCT releases the HI-TRAC. The IP-3 door is opened and HI-TRAC is positioned inside the IP-3 FSB truck bay beneath the overhead crane using the IP-3 air-pads and the site's tug vehicle. The IP-3 FSB truck bay door is closed. The HI-TRAC annulus is filled with water to a height of about 10 inches below the top of HI-TRAC.

This water helps minimize contaminated particulates from the outside of the STC, if any, from adhering to the inside of HI-TRAC. It also assures that when the loaded STC is placed back in to HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT Hl-2094289 1-12 Rev. I

the HI-TRAC the water does not overflow. The STC lid bolts are removed and if not already, the STC is filled with pool water.

The overhead crane is positioned over the STC and the STC lift rig is attached to the overhead crane. The lift rig connects the overhead crane to the STC lid and the STC is removed from the HI-TRAC and positioned over the cask loading area of the spent fuel pool. A set of remotely actuated connectors attach the STC lid to the STC body. The remotely actuated connectors can be operated manually if the actuator fails. The STC is lowered into the cask loading area and the lid is removed. The lid is removed from the spent fuel pool and the lid seal is inspected and replaced as necessary.

For each fuel transfer cycle, up to twelve spent fuel assemblies may be loaded into the STC. The STC lid is positioned over the STC and installed. The lid is installed and the remotely actuated connectors attach it to the STC body. The STC is then raised to the surface of the spent fuel pool and any standing water on the lid is removed. Several lid bolts are installed and tightened to hand tight. The water level in the STC is lowered using a lid drain penetration equipped with a fixed length tube. This ensures the correct amount of water is removed from the STC cavity and prevents over draining the cavity. In accordance with direction from the site's Radiation Protection Group the STC is raised and removed from the spent fuel pool, sprayed with demineralized water and placed directly into the HI-TRAC. The remaining STC lid bolts are installed and the lid bolts are tightened. The lift rig is disconnected from the STC top lid. HI-TRAC's top lid is installed and the bolts are tightened and the seal is tested in accordance with ANSI N14.5 [B.T]. The IP-3 door is opened and the HI-TRAC is moved to the Vertical VCT on air-pads.

The VCT will travel an approved route between IP-3 and IP-2. The load path will be evaluated (i.e. roadway and underground facilities) prior to the transfer of the spent fuel and upgraded as necessary to support the VCT with the loaded STC in the HI-TRAC. The evaluation, performed under 50.59, will consider a spent fuel transfer path starting at the IP-3 cask loading area and traveling to the IP-2 cask loading area. If any portion of the path has been analyzed previously as part of the IP-2 dry storage campaign, that analysis will be considered bounding since the VCT carrying a loaded MPC inside the HI-STORM weighs more than the VCT carrying a loaded STC inside the HI-TRAC. Prior to each transfer, the roadway will also be visually inspected and repaired as necessary.

Prior to the fuel transfer the boundary of the protected area will be changed so that the VCT will always remain within the protected area. The site security plan will also be modified as required.

The HI-TRAC containing the loaded STC is lowered from the VCT onto the IP-2 LPT and moved into the IP-2 FSB. Inside the IP-2 FSB, the HI-TRAC is positioned beneath the cask handling crane. A drain line containing a pressure gauge is connected to HI-TRAC's top lid penetration valve. The valve is slowly opened and any internal pressure is relieved. The HI-TRAC top lid bolts are removed and the HI-TRAC top lid is removed. The drain line is then attached to the vent port valve located on the lid of the STC. The valve is slowly opened and any HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-13 Rev. 1

internal STC pressure is relieved. STC lid bolts are loosened but left in place to secure the lid during STC lifting and handling.

The STC lift rig is attached to the STC lid. The cask handling crane is attached to the STC lift rig. In accordance with direction from the site's Radiation Protection Group the STC is raised and positioned directly into the spent fuel pool cask loading area and lowered so that the lid remains above the water.

Procedures are already in place to confirm acceptable boron levels in the IP-2 SFP every seven days per IP-2 TS 3.7.13 and at more frequent intervals when moving fuel assemblies. The minimum soluble boron concentrations of the IP-3 and IP-2 pools are 1000 ppm and 2000 ppm, respectively. A dilution analysis of the IP-2 SFP is required since the loaded STC will be filled with water from the IP-3 pool. A bounding analysis was done in accordance with NET-173-02

[T.K] assuming the entire STC cavity is completely filled with unborated water. The analysis does not consider the amount of water which will be displaced by the fuel assemblies or the amount of water which will be removed from the top of the cavity prior to transfer. The soluble boron level in the IP-2 SFP is reduced by less than 1% and is considerably higher than the required 786 ppm ensuring sub-criticality under non-accident conditions [T.L].

The lid bolts are removed and the STC is lowered completely into the cask loading area. The remotely-actuated connectors release from the STC body and the STC lid is removed. The STC lid is removed from the spent fuel pool and the spent fuel assemblies are removed from the STC.

The STC lid seal is inspected and replaced as necessary. The STC lid is positioned over the STC and installed. The lid's remotely-actuated connectors attach to the STC body and the STC is raised to the surface of the spent fuel pool. Any standing water on the lid is removed. Several lid bolts are installed and tightened to hand tight to provide a redundant lid attachment. In accordance with direction from the site's Radiation Protection Group the STC is raised and removed from the spent fuel pool, sprayed with clean water and placed directly into the HI-TRAC. The remaining STC lid bolts are installed and the lid bolts are tightened. The lift rig is disconnected from the STC top lid. The HI-TRAC top lid is installed and the bolts are tightened and the HI-TRAC containing the empty STC is then ready to be returned to IP-3.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-14 Rev. I

FIGURE 1.4.1:

A HI-TRAC CONTAINING AN EMPTY SHIELDED TRANSFER CANISTER IS POSITIONED IN THE IP-3 CASK RECEIVING BAY (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-15 Rev. I

FIGURE 1.4.2:

THE HI-TRAC TOP LID IS REMOVED AND THE SHIELDED TRANSFER CANISTER LID BOLTS ARE PREPARED (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL' REPORT HI-2094289 1-16 Rev. I

FIGURE 1.4.3:

THE OVERHEAD CRANE IS ATTACHED TO THE SHIELDED TRANSFER CANISTER TOP LID (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-17 Rev. I

FIGURE 1.4.4:

THE SHIELDED TRANSFER CANISTER IS REMOVED FROM HI-TRAC (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-18 Rev. I

FIGURE 1.4.5:

THE SHIELDED TRANSFER CANISTER IS POSITIONED IN THE SPENT FUEL POOL WITH THE LID ACCESSIBLE WHERE THE REMAINING INSTALLED LID BOLTS ARE REMOVED (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-19 Rev. 1

FIGURE 1.4.6:

THE SHIELDED TRANSFER CANISTER IS LOWERED INTO THE CASK LOADING AREA OF THE SPENT FUEL POOL (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT Hl-2094289, 1-20 Rev. I

FIGURE 1.4.7:

THE SHIELDED TRANSFER CANISTER LID IS REMOVED (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-21 Rev. I

FIGURE 1.4.8:

SPENT FUEL ASSEMBLIES ARE LOADED INTO THE SHIELDED TRANSFER CANISTER (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-22 Rev. I

FIGURE 1.4.9:

THE SHIELDED TRANSFER CANISTER IS FULLY LOADED WITH SPENT FUEL (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-23 Rev. 1

FIGURE 1.4.10:

THE SHIELDED TRANSFER CANISTER LID IS INSTALLED UNDERWATER (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-24 Rev. 1

FIGURE 1.4.11:

THE SHIELDED TRANSFER CANISTER IS RAISED TO THE TOP OF THE SPENT FUEL POOL (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-25 Rev. I

FIGURE 1.4.12:

A SMALL VOLUME OF WATER IS PUMPED FROM THE SHIELDED TRANSFER CANISTER CAVITY (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-26 Rev. I

FIGURE 1.4.13:

THE SHIELDED TRANSFER CANISTER IS REMOVED FROM THE SPENT FUEL POOL AND PLACED DIRECTLY IN HI-TRAC (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-27 Rev. I

FIGURE 1.4.14:

THE SHIELDED TRANSFER CANISTER IS FULLY SEATED IN HI-TRAC AND THE LID BOLTS ARE TIGHTENED (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-28 Rev. I

FIGURE 1.4.15:

THE HI-TRAC TOP LID IS INSTALLED AND THE BOLTS ARE TIGHTENED (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-29 Rev. I

FIGURE 1.4.16:

HI-TRAC IS MOVED FROM THE IP-3 CASK RECEIVING AREA TO THE AWAITING VERTICAL CASK TRANSPORTER (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-30 Rev. I

FIGURE 1.4.17:

THE VERTICAL CASK TRANSPORTER LIFTS THE HI-TRAC FROM THE AIR-PADS AND MOVES IT TO IP-2 (FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-31 Rev. I

1.5 Reference Drawings The following drawings are provided on subsequent pages in this section:

DRAWINGS 6013, 6015 and 6571 are withheld in accordance with 10 CFR 2.390.

Drawing Title Revision No.

6013 INDIAN POINT UNIT 3 SHIELDED TRANSFER CANISTER Rev 6 ASSEMBLY 6015 INDIAN POINT UNIT 3 SHIELDED TRANSFER CANISTER Rev 3 BASKET ASSEMBLY 6571 INDIAN POINT UNIT 3 HI-TRAC TRANSFER CASK TOP Rev 1 LID ASSEMBLY 4128 HI-TRAC 100D TRANSFER CASK Rev 6 The current licensing drawing for the HI-TRAC lOOD is attached (Holtec Drawing No. 4128 Revision 6) for informational purposes only.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-32 Rev. I

8 7 6 5 -JL 1 CLIENT GENERAL LICENSING DRAWING PACKAGE COVER SHEET PROJECT NO. 1026 P.O. NO. N/A DRAWING 418 TOTAL 1 D PACKAGE I.D. SHEETS 10 REVISION LOG REV AFFECTED DRAWING

SUMMARY

OF CHANGES/ PREPARED APPROVAL MIR#

SHEET NUMBERS AFFECTED ECOs BY DATE 0 INITIALISSUE 1026-30 S.CAIN 8/25/03 82748 LICENSING DRAWING PACKAGE CONTENTS: .1 ALL SHEETS 1026-31 T.F.O. 10127103 70107 2 SHEET4 1026-32 LEH 12/17/03 86122 SHEET DESCRIPTION 3 SHEET 5 &8 1026-33 T.F.O. 3123/04 30678 1 COVER SHEET 4 SHEET 6 1026A-4 JJB 12/30/05 62382 2 ASSEMBLY DRAWINGAND BILLOF MATERIALS 5 SHEET4 1026-41 SLC 21l5/06 61656 3 OVERALLDIMENSIONS 4 POOL LID ASSEMBLY 6 SHEET B 1026-42 SLC 8/12/08 57791 5 BASE PLATE ASSEMBLY C

6 OUTER SHELL ASSEMBLY 7 TOP FLANGE DETAILS 8 TRUNNIONAND INNERSHELL ASSEMBLY 9 WATER JACKET SHELL ASSEMBLY THE VALIDATION IDENTIFICATION 1VIR)NUMBER RECORD ISACOMPUTER GENERATED NUMBER RANDOM WHICH 10 TOP LIDASSEMBLY CONFIRMSTHATALLAPPROPRIATE REVIEWSOFT*IS DRAWING AREDOCUMENTED INCOMPANYSNETWORK.

k.

U-NOTES:

1. THE EQUIPMENT DOCUMENTEDINTHIS DRAWINGPACKAGE HAS BEEN CONFIRMED BY HOLTEC INTERNATIONAL TO COMPLY WITHTHE SAFETY

- I ANALYSES DESCRIBED IN THE HI-STORM FSAR.

2. DIMENSIONALTOLERANCES ON THIS DRAWINGARE PROVIDED SOLELYFOR R LICENSING PURPOSES TO DEFINE REASONABLE LIMITSON THE NOMINAL DIMENSIONS USED IN LICENSINGWORK. HARDWAREIS FABRICATEDIN ACCORDANCE WITH THE DESIGN DRAWINGS, WHICHHAVEMORE RESTRICTIVE -

TOLERANCES, TO ENSURE COMPONENT FIT-UP. DO NOT USE WORST-CASE TOLERANCE STACK-UP FROM THIS DRAWINGTO DETERMINE COMPONENT FIT-UP.

I 3. THE REVISION LEVEL OF EACH INDIVIDUALSHEET IN THE PACKAGE IS THE SAME AS THE REVISION LEVELOF THIS COVER SHEET. A REVISION TO ANY SHEET S IN THIS PACKAGE REQUIRES UPDATING OF REVISION NUMBERS OF ALL SHEETS TO THE NEXT REVISION NUMBER.

4. APPLICABLECODES AND STANDARDSARE DELINEATEDIN FSAR SECTION 2.2A.

5. ALL WELDS REQUIRE VISUALEXAMINATION.ADDITIONALNDE INSPECTIONS ARE NOTED ON THE DRAWING.NOE TECHNIQUES AND ACCEPTANCE CRITERIA ARE PROVIDED IN FSAR TABLE9.1.4.

I

6. UNLESS OTHEWISE NOTED, FULL PENETRATION WELDS MAY BE MADE FROM EITHER SIDE OF A COMPONENT.

ISOMETRIC HI-TRACIEWOF 100D

7. THIS COMPONENT IS IMPORTANT-TO-SAFElY, CATEGORY A, BASED ON THE HIGHEST TRANSFER CAS CLASSIFICATIONOF ANY SUBCOMPONENT. SUBCOMPONENT CLASSIFICATIONS ARE PROVIDED ON THE DESIGN DRAWING.

"NOTICECERTAIN EQUIPMENT DESIGN AND AREPROTECTED OPERATIONAL UNDER FERATU U.S.PATENT RES NO. OFTH.S 6,5M7,536.

8. ALL WELD SIZES ARE MINIMUMSEXCEPT AS ALLOWED BY APPLICABLECODES AS ANINFRINGEMENT BYANYENTITYIS UNLWFUL."

CLARIFIEDIN THE FSAR. FABRICATOR MAYADD WELDS WITH HOLTEC APPROVAL. A

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"**=*' H OL T EC 100TON INTE HI-TRAC 100D ASSEMBLY 1026 4i428 3I N/A NOE I 8 7 ~6543 87 5 4 3 1

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EXPLODED ISOMETRIC VIEW POOL LID LEAD SHIELD POOL LID RING C

--078 W'2REF. -_i_ T .1/2' B

POOL LID BOTTOM 3130" REF POOL LIDTOP "NOTICE:

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BYANYENTITY ANINFRINGEMIENT ISUNLAWFUL" 12 TYP . A Fl /I F\n fflfllflGENERAL REENOO I HOLTEC

. INTERNATIONAL 100TON HI-TRAC-100D SECTION E-E 5~ ....

... -TR11'AR BASE PLATE ASSEMBLY W 1 aN SCALE BASE PLATE WELDMENT 1026

'j128 7B5,3 8 7 6 5 t, 3 2 1 I CR35 _____________________ 11 I -

8 7 5 4 3 2 1 T '"M ASSEMBLY ISOMETRIC VIEW u WATER JACKET TOP PLATE Fý SECTION

'VT& PT ORM? OFThIS NOTICE:CERTAIN ANDOPERATIONAL DESIGN FEATURES EQUIPMENT UNDER AREPROTECTED US. PATENTNO,6,587,536 ANINFRINGEMENTBYANYENTITY IS UNLAWFUL."

EMEEMF GENERAL HOLTEC

lOOTON INTERNATIONAL HI-TRAC 100D OUTER SHELL ASSEMBLY OUTER SHELL WELDMENT 4128 6 5 1 8 7 4 21 876543

ASSEMBLY ISOMETRIC VIEW" G,(REF.)

TOP FLANGE NOTICE:CERTAIN DESIGNANDOPERATIONAL FEATURESOFTHIS EDUIPMENT AREPROTECTEDUNDER US. PATENTNO.6,57,.536.

ANINFRINGEMENTBYANYENTITYISUNL.AFUL,."

I i I SENSE GENERAL HOLTEC 100 TON INTERNATIONAL HI-TRAC 100D a,*ME"oE TOP FLANGE ASSEMBLY

-- *1-" 4128 I' 1 I NON

ASSEMBLY ISOMETRIC VIEW 7 REF, 0 VT&PT I, 5"P SH" TRUNNION BLOCK LIFTING TRUNNION 60 INNER SHELL DETAIL N

L -i

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CERTAIN DESIGN ANDOPERATIONAL FEATURESOF THIS EQUIPMENT AREPROTECTEDUNDER U.S.PATENTNO.6,587.53.5 ANINFRINGEMENT BYANYENTITY IS U*LAWFUL.

LI>D EMENEM GENERAL 3W8 VT&PT HOLTEC INTERNATIONAL 1OOTON 8 TYP. HI-TRAC 100D SECTION KR TRUNNION & SHELL ASSEMBLY SECTION"-

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"NOTCEE CERTAINDESIGN ANDOPERATIONALFEATURES OF THIS EOQUPN4TAREPROTECTEDUNDERUS. PATENTNO.6,5R7,53R.

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1.6 Supplier's Qualification The inter-unit transfer program is being managed by Holtec International as the prime contractor to Entergy. Holtec International is an engineering technology company with a principal focus on the power industry. Holtec's Nuclear Power Division (NPD) specializes in spent fuel storage technologies, both wet and dry.

The USNRC dockets in parts 71 and 72 currently maintained by the Company are listed in Table 1.6.1. Holtec's corporate engineering consists of professional engineers and experts with extensive experience in every discipline germane to the fuel storage technologies, namely structural mechanics, heat transfer, computational fluid dynamics, and nuclear physics.

Holtec International's quality assurance program was originally developed to meet NRC requirements delineated in 10CFR50, Appendix B, and was expanded to include provisions of 1 OCFR7 1, Subpart H, and 10CFR72, Subpart G, for structures, systems, and components designated as important to safety. The Holtec quality assurance program, which satisfies all 18 criteria in 10CFR72, Subpart G, that apply to the design, fabrication, construction, testing, operation, modification, and decommissioning of structures, systems, and components important to safety is incorporated by reference into this licensing report. Holtec's QA Program has been certified by the USNRC (Certificate No. 71-0784).

The equipment required for the inter-unit transfer project will be fabricated by Holtec Manufacturing Division (HMD) located in Pittsburgh, Pennsylvania. HMD is a long term N-Stamp holder and fabricator of nuclear components. Both Holtec's headquarters and the HMD subsidiary have been subject to triennial inspections by the USNRC. Although unlikely, if another fabricator is to be used for the fabrication of any equipment in this program, then the proposed fabricator will be evaluated and audited in accordance with Holtec International's quality assurance program.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-33 Rev. I

TABLE 1.6.1 USNRC DOCKETS ASSIGNED TO HOLTEC INTERNATIONAL System Name Docket Number HI-STORM 100 (Storage) 72-1014 HI-STAR 100 ( Storage) 72-1008 HI-STAR 100 (Transport) 71-9261 HI-STAR 180 (Transport) 71-9325 HI-STAR 60 (Transport) 71-9336 Holtec Quality Assurance Program 71-0784 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1-34 Rev. 1

CHAPTER 2: FUEL ACCEPTANCE CRITERIA AND ENGINEERED MEASURES FOR SAFETY 2.1 Fuel Acceptance Criteria The principal design parameters of the IP-3 fuel are provided in Table 4.5.1. This fuel is essentially identical to the IP-2 fuel. In order to be eligible for inter-unit transfer in the STC in accordance with the proposed Technical Specifications and analysis presented in this report, the IP-3 SNF must fulfill the following criteria:

i. The fuel must be intact as defined in the glossary.

ii. The initial average assembly enrichment must be less than 5 wt% U 23 5 . (Criticality) iii. It must have bumup less than 55,000 MWD/MTU. (ALARA and Thermal) iv. It must have burnup greater than 40,000 MWD/MTU or have a minimum burnup as a function of the initial enrichment shown in Table 4.7.3 to be placed in any location in the STC (Criticality).

v. Fuel not meeting the minimum burnup requirements may only be loaded in the outer or peripheral cells of the STC and the inner or central cells must remain empty. (Criticality) vi. It must have a minimum cooling time of 5 years. (ALARA and Thermal) vii. It must meet the decay heat limits given in Chapter 5 for the location in which it is to be loaded. (Thermal) viii. It may or may not contain non-fuel hardware.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 2-1 Rev. I

2.2 Safety and Protective Measures 2.2.1 Criticality Safety through Physical Design The acceptance criteria given in Section 2.1 ensure that the reactivity criteria set forth in Chapter 3 will be met by the SNF that will be loaded in the Shielded Transfer Canister (STC). The above statement is based on comparing the design data that directly affects reactivity of the spent fuel storage devices in the IP-2 SFP (TS Section 4.3) and the IP-3 SFP (TS Section 4.3) and the STC (Reference Drawing 6015) as compiled in Table 2.2.1. The following observations provide the basis for concluding that criticality safety is assured.

i. The areal B-10 density in the STC fuel basket is substantially greater than that in IP-2 and IP-3 Region 2 racks. A greater B- 10 loading corresponds to reduced reactivity.

ii. The thickness of the stainless steel walls in the STC fuel basket is considerably greater than that in the fuel racks in either pool. An increased mass of stainless steel reduces reactivity.

The above comparisons lead to the conclusion that the same criticality safety of the spent fuel stored in either pool is automatically assured in the STC fuel basket.

2.2.2 Criticality Safety through Assured Boron Concentration The fuel transferred to the STC is surrounded by its native environment in the pool, which is the pool's borated water. After the STC is raised from the pool, the boron concentration in the STC cavity will be the same as the pool and cannot be reduced by dilution or any other means. Thus, unlike storage in the fuel pool where the administrative controls are necessary to maintain boron concentration specified in the TS, the STC will, by virtue of its sealed configuration, maintain the boron concentration throughout the transfer process. The assured presence of soluble boron in the STC cavity adds another layer of safety against violation of the postulated reactivity limit in Chapter 3.

2.2.3 Release Protection by Multiple Barriers The proposed inter-unit transfer operation incorporates three independentbarriers against release of radioactivity to the environment, namely:

i. The fuel cladding (only intact fuel is permitted to be transferred) ii. The pressure tested STC is qualified to withstand a normal internal pressure of 50 psig (see Chapter 3).

iii. The HI-TRAC 100)D Transfer Cask is qualified to withstand a normal internal pressure of 30 psig.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 2-2 Rev. I

The number of barriers against release of radiological matter to the environment during the transfer process between the two fuel buildings, therefore, exceeds that present in wet storage in any fuel pool.

2.2.4 Protection by a Favorable Thermal-Hydraulic Environment As shown in Chapter 5, the thermal-hydraulic environment around the spent nuclear fuel in the STC basket is considerably more benign than that in the reactor vessel (where the peak water temperature is approximately 600'F or 315'C). The bounding fuel cladding temperature in the STC, presented in Table 5.4.1, is significantly lower by -360 'F or 200 'C. The empirical Arrhenius rule used to estimate the rate of chemical action (attack) on metals holds that the rate of reaction doubles with every 10'C rise in the aqueous temperature. Therefore, the thermal environment to which the fuel is subjected in the reactor is significantly more aggressive than in the STC's. Therefore, the risk of degradation of the fuel cladding during the transfer operation is ruled out.

2.2.5 Protection by the Selection of Low Dose Emitting Fuel From the population of fuel in the IP-3 pool, the specific SNFs selected for the inter-unit transfer shall have achieved a sufficient decay time so as to meet the heat load limit in Chapter 5. As the radiation emitted by the fuel decreases exponentially with the passage of time, the batch selected for transfer will have a correspondingly low dose accretion rate. This is borne out by the shielding analysis summarized in Chapter 7 where it is shown that the effect on the site boundary dose of a loaded HI-TRAC 100D when it is outside the Part 50 structure is negligible.

2.2.6 Protection by Use of Proven Equipment The HI-TRAC 100)D Transfer Cask which serves as a principal radiation barrier in the inter-unit transfer operations is a proven piece of equipment through multiple uses in the IP-2 dry storage campaign mentioned in Section 1.1. The radiation shielding capacity of the HI-TRAC 100D is described in the FSAR [K.A]. It is part of the HI-STORM 100 System approved by the USNRC and demonstrated by measurement. Therefore, the safety of the inter-unit transfer operation is ensured to be ALARA.

2.2.7 Protection by Material Selection The STC and HI-TRAC 100)D are two principal components whose materials of construction must be assured from an adverse performance.

As discussed in Chapter 8, the materials used in the manufacture of the STC are of the same genre as used in the fabrication of casks and fuel baskets in Holtec's dry storage program. The suitability of these materials, including surface preservatives, has been endorsed by the USNRC on several active Holtec dockets. Therefore, the risk of an anomalous performance by an STC material is unlikely.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 2-3 Rev. 1

The HI-TRAC IOOD Transfer Cask is, as mentioned above, a proven hardware having been used in a virtually identical environment in the IP- 1 and IP-2 dry storage campaign. Therefore, the risk of a material malfunction during the inter-unit transfer campaigns is unlikely.

2.2.8 Reliability through Increased Structural Margins As summarized in Chapter 6 herein, the SSCs proposed for use in the inter-unit transfer program have been engineered with significantly larger structural margins of safety than required to meet the applicable design criteria.

Specifically:

i. The STC is engineered to maintain the stress levels in its pressure boundary to well below the Code allowables.

ii. The tensile strength of the flange bolts in the STC is considerably larger than that required to maintain the joint seals.

iii. The HI-TRAC 100D will maintain its stress levels when subject to the Design Pressure values that are considerably lower than the respective Code allowables.

iv. Special lifting devices such as the lift yoke, the lift cleat and other lifting appurtenances are designed to meet the stress limits of ANSI N14.6 and NUREG-0612, Section 5.1.6.(1)(a) with ample margins. These special lifting devices will be load tested in accordance with ANSI N 14.6 prior to use. Other lifting interfaces such as the trunnions, the STC and HI-TRAC lid lifting points are designed per guidance from NUREG-0612, Section 5.1.6 (3).

v. The IP-2 and IP-3 cask handling cranes will be single-failure-proof and comply with NUREG-0554 and NOG-1-2004. The vertical cask transporter also has a redundant drop protection feature.

The above attributes of the STC, HI-TRAC IOOD and the lifting equipment ensure that the SSCs involved in the inter-unit transfer operation shall not suffer from a structural malfunction or failure.

2.2.9 Protection by Design to Prevent Inadvertent Water Loss The STC is welded cylindrical layered (steel-lead-steel) canister with a welded steel base plate.

The top lid is bolted and sealed using an elastomeric seal which will be pressure tested to ensure no leakage of water can occur. There are no penetrations in the STC which allow a release of water after it is sealed. During operations, prior to the lid being sealed, water levels will be monitored and maintained at the required levels. There is no malfunction or accident which would cause a loss of water from the STC cavity.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 2-4 Rev. I

The HI-TRAC is a welded cylindrical vessel with a bolted top lid, containing an elastomeric seal, and a bolted pool (bottom lid), containing an elastomeric seal and pipe plug in Pool Lid drain.

These seals and plug will be pressure tested to ensure no leakage of water can occur. There are no other penetrations which will allow a release of water after it is sealed. During operations, the water levels in the HI-TRAC will be monitored and maintained at the required levels. There is no malfunction or accident which would cause a loss of water from the HI-TRAC cavity.

Table 2.2.1 COMPARISON OF REACTIVITY-INFLUENTIAL DATA Item STC Fuel IP-2 Region 2-2 IP-3 Region 2 Basket Racks Racks

1. B-10 loading, g/cm 2 0.032 0.026 0.020

2. Nominal thickness of 0.281 0.075 0.085 stainless steel cell walls, in. I I HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 2-5 Rev. 1

Chapter 3: Principal Design Criteria, Applicable Loads, and Postulated Accidents 3.1. Governing Regulatory Requirements 3.1.1. Criticality The fuel basket in the STC, as described in Section 1.3, consists of a honeycomb assemblage of prismatic cells with one panel of the Metamic neutron absorber attached to each stainless steel cell wall. From the reactivity control standpoint, the fuel basket simulates the Region 2 storage racks in IP-3, albeit with a larger cell-to-cell pitch and a greater B-10 loading. (The IP-2 racks were supplied by Holtec International). The IP-3 racks were provided by a supplier, now owned by Holtec International. The fuel transferred to the IP-2 pool is required to have accumulated the burnup so that it can be storable in both IP-2 and IP-3 pools. Therefore, the required minimum burnup as a function of enrichment specified in Table 4.7.3 is the more limiting of IP-3 Region 2 and IP-2 Region 2-2 racks. IP3 fuel will be stored at IP-2 in accordance with IP-2 TS 3.7.13.

The following criticality safety requirements apply:

i. The effective neutron multiplication factor (kff) shall be less than 0.95 with the STC fully loaded with fuel of the highest anticipated reactivity and the STC cavity flooded with unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity shall include a margin for uncertainty in reactivity calculations including manufacturing tolerances and shall be shown to be less than 0.95 with a 95% probability at a 95% confidence level.

ii. Reactivity effects of abnormal and accident conditions shall be evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95, with credit for soluble boron to offset the accident condition.

Applicable codes, standard, and regulations or pertinent sections thereof, include the following:

Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62, "Prevention of Criticality in Fuel Storage and Handling."

USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Rev. 3 - March 2007.

o USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications (GL-78-011), including modification letter dated January 18, 1979 (GL-79-004).

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-1 Rev. 1

L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L.

Kopp to T. Collins, August 19, 1998.

" USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2, March 2007.

ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

" Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements" 3.1.2. Shielding The STC provides shielding to maintain occupational exposures ALARA in accordance with 10CFR20, while also maintaining the maximum load on IP-3's crane hook to below the rated capacity of the crane. The calculated dose rates around the loaded STC are reported in Chapter 7. The calculated dose rates around the loaded HI-TRAC are also reported in Chapter 7. These dose rates are used to perform an occupational exposure estimate for the transfer operations. A postulated HI-TRAC accident condition where there is a loss of water in the water jacket (i.e. the neutron shield), is also evaluated in Chapter 7.

The shielding criteria are derived from 10 CFR Part 20, 10 CFR Part 100, 10CFR72.104 and 10CFR72.106 which provide radiation dose limits for any real individual located at or beyond the nearest boundary of the controlled area. The acceptance criteria from 10 CFR 72 were used rather than 10 CFR 100, since the 10 CFR 72 regulations are more restrictive. The individual must not receive doses in excess of the limits given in Chapter 7 for normal, off-normal, and accident conditions.

The objective of shielding is to assure that radiation dose rates at key locations are as low as practical in order to maintain occupational doses to operating personnel As Low As Reasonably Achievable (ALARA) and to meet the requirements of 10CFR 72.104 and 10CFR 72.106 for dose at the controlled area boundary. Three locations are of particular interest during the inter-unit transfer operations are:

immediate vicinity of the STC and HI-TRAC protected area boundary owner controlled area (site) boundary HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-2 Rev. 1

Dose rates in the immediate vicinity of the loaded STC and HI-TRAC are important in consideration of occupational exposure. Conservative evaluations of dose rate have been performed and are described in Chapter 7.

There are no expected radioactive effluents that result from STC or transfer operations using HI-TRAC 100D, nevertheless this is addressed in Chapter 7.

Detailed operating procedures for the inter-unit transfer shall be prepared based on Chapter 10, and site-specific requirements, including the Part 50 Technical Specification of IP-2 and IP-3 in keeping with ALARA.

As discussed in Chapter 7, ALARA principles in accordance with 10CFR20.1 101(b) shall be applied to keep the dose rates and personnel exposures to the lowest possible values.

3.1.3. Thermal The STC thermal wet transfer evaluation acceptance criteria are stated below:

The spent fuel cladding temperatures must be below 400 'C in accordance with SFST-ISG- 11 Rev.3 [E.K] for short-term operations since no similar regulations/limits exist in 10 CFR Part 50.

" The STC and the HI-TRAC components must remain below the HI-TRAC temperature limits as specified in the HI-STORM 100 FSAR [K.A].

The applicable temperature limits are summarized from the above sources in Table 3.1.1.

The specified temperature limits in Table 3.1.1 are either equal to or less than those in

[K.A] for conservatism.

Finally, the total heat load transferred to the IP-2 fuel pool by the inter-unit transfer in any campaign must be limited by the design basis heat load specified in the IP-2 UFSAR.

This limitation is addressed in Chapter 5 of this report.

3.1.4. Structural 3.1.4.1 Overview The structural qualification of the components used in the inter-unit transfer must consider both normal and accident conditions. Normal condition, as the term implies, is the expected bounding condition that will prevail during the transfer operation. An accident condition is a hypothetical, yet statistically credible event that may have an adverse effect on an SSC or the transfer operation. The object of the criteria is to ensure that the margins of safety under all postulated accidents will remain sufficiently large to preclude any regulatory safety concerns. The SSCs whose safety must be evaluated to support the safety case for the inter-unit transfer are:

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-3 Rev. I

i. The Shielded STC (STC) ii. The Fuel Basket inside the STC iii. The HI-TRAC IOOD Transfer Cask The STC and its fuel basket are new components whose design criteria are described in this section. HI-TRAC 1OOD, however, is an NRC-certified cask (Docket No. 72-1014) whose design basis is fully articulated in the HI-STORM 100 FSAR [K.A]. However, the HI-TRAC's ability to withstand all normal and accident condition loads applicable to the inter-unit transfer operation are identified and evaluated in this Licensing Report.

During short-term operations, the STC is designed to meet the Level A stress limits of the ASME Code,Section III, Subsection ND (2004 Edition). In addition, the components of the single-failure-proof lifting system used to handle the STC and HI-TRAC will meet the guidance in Section 5.1.6 of NUREG-0612 and in ANSI N14.6, as applicable. The IP-3 crane is being upgraded to be single failure proof meeting the requirements of NUREG-0554. Since both the IP-3 and IP-2 cask handling cranes will be single failure proof, a drop accident involving the STC inside the FSB is not credible. The STC will be contained in and carried along the haul path in the HI-TRAC using the site's Vertical Cask Transporter (VCT). The STC weighs less than a loaded MPC (set at 45 tons in the HI-STORM 100 FSAR); therefore the structural analysis of the IP-2 VCT loaded with the STC is bounded by the analysis performed under Part 72 where the HI-TRAC is loaded with 32 fuel assemblies in an MPC.

The most severe load on the fuel basket inside the STC is a fuel handling accident resulting in a free drop of a fuel assembly onto the top of the fuel basket. The acceptance criterion for this accident event is that the fuel storage array remains subcritical. In other words, the plastic deformation of the fuel basket cell wall due to the fuel impact must not extend down into the active fuel region.

In Docket No. 72-1014, the HI-TRAC IOOD is designed to meet the stress limits of the ASME Code,Section III, Subsection NF (1995 Edition), Class 3. It is also noted that the HI-TRAC lOOD, with the new bolted top lid, will be subject to an internal pressure when it is used to transport the STC. The HI-STORM FSAR does not consider any internal pressure loads on the HI-TRAC 100D since the lid used for dry storage contains a large circular hole in the middle. Therefore, the newly designed top lid, as well as the HI-TRAC inner shell and bottom lid, must be evaluated for the effects of internal pressure.

For this purpose the stress limits of ASME Section III Subsection ND (pressure vessel code) shall be used.

In the next section, all loadings germane to normal and accident conditions are compiled along with the associated acceptance criteria.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-4 Rev. 1

Table 3.1.1 TEMPERATURE LIMITS APPLICABLE TO INTER-UNIT TRANSFER Item Condition Short-Term Abnormal or accident Operations condition (Normal)

1. Fuel Cladding 400°C (752°F) 570 0 C (1058-F)

2. STC Metal Parts 150-C (302°F) 200-C (392 0F)

3. HI-TRAC SealsNo e t 3 120-C (248 F)0 170TC (338°F) up to 20 hrs 210TC (410 0 F) up to 3 hrs

4. t STC SealNo e 3 120-C (248°F) 120-C (248°F)

5. HI-TRAC Metal Parts 150-C (302°F) 2000 C (392 0F)

6. HI-TRAC Water Jacket 153 0 C (307 0 F) Ote N/ANOte

7. HI-TRAC annulus water 0

134 C (274 0 F)No*e I 142 0C (287 0 F) Note 1 1480 C (298 0 F)Note I 155.5 0 C (312 F)Note I 0

8. STC Water Note 1: The bulk temperature of water must be limited to the boiling temperature of water at the enclosure design pressure.

Note 2: The water jacket is equipped with safety relief devices to prevent over-pressure during accidents.

Note 3: The temperature rating of seals must meet or exceed the requirements specified herein.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-5 Rev. 1

3.2. Applicable (Design, Normal, and Postulated Accident) Loadings 3.2.1. Design Basis Loads The design pressure and associated design temperature for the STC and HI-TRAC IOOD are provided in Table 3.2.1. Both components are required to meet the applicable stress limits under the provisions of ASME Section III Subsection ND, Class 3.

The analysis of the normal and various accident condition loads must show that the internal pressure and temperature remain below their respective limits in Table 3.2.1.

3.2.2. Normal Condition Loads The operating pressure and temperature under normal operations, presented in Tables 5.3.2 and 5.3.1, respectively; are bounded by the design pressure and temperature in Table 3.2.1 and, therefore, do not require a separate analysis.

3.2.3. Accident Condition Loads Accident conditions for inter-unit operations belong to two categories, namely:

i. Credible accidents inside Part 50 structure ii. Credible and non-credible accidents during the transfer of loaded HI-TRAC 1OOD from IP-3 to IP-2 Fuel Building 3.2.3.1 Accidents Inside Part 50 Structure The accident conditions postulated are summarized below.

(a) Accidental drop of a fuel assembly: As discussed previously, all heavy load handling evolutions will be performed using single-failure-proof lifting systems.

Therefore, an accidental drop of a heavy load inside Part 50 structure is not credible. However, this assertion does not apply to individual spent nuclear fuel (SNF) assemblies which are handled using a tool that does not have redundant drop protection features against an accidental drop. Therefore, the scenario of an accidental drop of an SNF on the fuel basket must be evaluated.

As noted in Generic Letter 78-11, O.T. Position Paper [F.A], a fuel assembly, along with the portion of the handling tool, which is severable in the case of a single element failure, is assumed to drop vertically and hit the top of the fuel basket. Inasmuch as the fuel basket is of honeycomb construction, the HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-6 Rev. 1

deformation produced by the impact is expected to be confined to the region of collision. However, the "depth" of damage to the affected cell walls must be demonstrated to remain limited to the portion of the cell above the top of the "active fuel region", which is essentially the elevation of the top of the neutron absorber.

Stated in qualitative terms, this criterion implies that the plastic deformation of the fuel box cell walls should not extend beyond the permissible value listed in Table 3.2.2. In order to utilize an upper bound of kinetic energy at impact, the impactor (fuel assembly including the handling tools) shall use bounding weight and height from Table 3.2.2.

Any radioactive release from the drop of the SNF is already analyzed and is presented in the fuel handling dose analysis in Chapter 14 of the UFSAR. This analysis bounds the accidental drop of a fuel assembly being loaded into the STC because the minimum cooling times of the fuel to be transferred to IP-2 (5 years) is significantly longer than the cooling time of the SNF analyzed (84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months ).

(b) Misloading of a fuel assembly into the STC basket: Two different scenarios are addressed to ensure both criticality safety and thermal performance of the STC.

Robust administrative controls as discussed in Chapter 10 will be in effect to ensure that either misload condition will not occur. It should be noted that the same procedures and controls are followed as part the fuel selection process for dry cask storage. This verification is performed independently by fuel handing and Reactor Engineering personnel. In addition to these administrative controls, operational controls will also be in place.

(i) For criticality (see Chapter 4), the misloading of a fuel assembly which has not attained required bumup for loading under Configuration 1, as well as the misloading of any fuel assembly into one of the four center STC basket cells in Configuration 2 is addressed. To prevent a misloaded assembly in Configuration 2, a cell blocker device can be used to block the openings of the four center cells during the loading of the STC. To prevent a misloaded assembly in Configuration 1, this same device will be used when loading the first eight assemblies and additional visual inspections can be implemented prior to and after loading each of the remaining four assemblies, as described in Chapter 10.

(ii) For thermal performance the misloading of a recently irradiated fuel assembly is considered. The misloading of a recently irradiated fuel assembly will cause an immediate temperature rise in the STC basket cell. Thermocouples will be placed and monitored such that if a recently irradiated fuel assembly is placed into the STC it will be HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-7 Rev. 1

immediately detected and removed prior to the fuel transfer. This is further described in Chapter 10.

(c) Earthquake: The stability of the loaded STC in the spent fuel pool under the site's Design Basis Earthquake (DBE) must be demonstrated by analysis. The site's DBE loads are provided in Table 3.2.2.

3.2.3.2 Accident Scenarios Outside Part 50 Structure Most accident scenarios during the brief period (normal duration of transfer will most likely occur in one work shift) when the VCT hauls the HI-TRAC from the IP-3 to the IP-2 truck bay thresholds are related to environmental phenomena.

Two postulated events not considered environmental are a postulated drop event which can occur when the HI-TRAC is being lifted or lowered and is not securely attached with the redundant drop protection pins to the VCT and a scenario where the duration of the transfer is extended as a result of a postulated VCT breakdown. All of these postulated events are summarized below.

(a) Accidental drop of loaded HI-TRAC 1OOD: The maximum height which the loaded HI-TRAC IOOD can be lifted is limited to the value in Table 3.2.2. The lift height of the loaded HI-TRAC will be controlled as it is raised on the VCT to ensure this limit is not exceeded. Once the locking pins are engaged, attaching the HI-TRAC to the VCT and providing redundant drop protection, the lift height/carry is no longer limited.

(b) Fire: The potential of a fire accident near the VCT during its movement is considered to be extremely remote because there are no significant combustible materials in the area. An evaluation of the haul path will be performed to ensure the fire accident considered in this licensing report bounds any site specific scenario. Transitory hazards will be controlled with administrative procedures.

The HI-TRAC 1OOD transfer cask fire accident is conservatively postulated to be the result of the spillage and ignition of 50 gallons of combustible fuel which engulfs the HI-TRAC. The HI-TRAC transfer cask surfaces are considered to receive an incident radiation and forced convection heat flux from the fire.

Table 3.2.3 provides the fire durations for the HI-TRAC 1OOD based on the amount of flammable materials assumed. The temperature of fire is assumed to be 1475°F to accord with the provisions of 10CFR71.73 since no guidance is supplied in Part 50.

(c) Lightning: The effect of a lightning strike on the transfer cask is considered in the Entergy HI-STORM 100 Cask System 72.212 Evaluation Report, IPEC Site Specific Appendix F [U.B] where it is determined that lightning will not impair the safety function of the cask. Lightning may however cause an ignition HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 1 3-8 Rev. 1

of the transporter fuel. That scenario is considered above. Therefore, this loading is not considered further in this Licensing Report.

(d) Earthquake: The stability of the loaded VCT and the loaded HI-TRAC standing alone under the site's Design Basis Earthquake (DBE) will be demonstrated by analysis. The site's DBE loads are provided in Table 3.2.2.

(e) Flood: Potential sources for the flood water could be unusually high water from a river or stream, a dam break, or a hurricane. The plant UFSAR states "Flooding at the site has been nonexistent. The highest recorded water elevation at the site was 7.4 feet above mean sea level during an exceptionally severe hurricane in November, 1950. Since the river water elevation would have to reach 15 feet 3 inches above mean sea level before it would seep into the lowest floor elevation of any of the Indian Point buildings, the potential for any flooding damage at the site appears to be extremely remote." The IP-2 and IP-3 FSB truck bay and transport haul path are well above the lowest building elevation and would require a rise in the river of over 55 feet to cause flooding; therefore the affect of the flood on the VCT is not considered credible and is not specifically analyzed.

(f) Environmental Loadings: The loadings from an extreme environmental phenomena, such as high winds, tornado, and tornado-borne missiles, as specified for the 48 contiguous states in Reg. Guide 1.76, ANSI 57.9, and ASCE 7-88, are considered in the certification of HI-TRAC IOOD in Docket No. 72-1014. These loadings bound the environmental loadings at IPEC.

Therefore, a site-specific analysis for the inter-unit transfer operation is not required.

(g) Loss of Water in the Water Jacket: As a conservative measure, the water in the water jacket of HI-TRAC IOOD is assumed to be lost. The resulting increase in the site boundary dose must be quantified to demonstrate compliance with the specified annual site boundary dose limit. Pressure and temperature limits must also be assured.

(h) Extended time of STC residence in the HI-TRAC: This accident condition postulates that, for whatever reason, the STC is kept in the transfer cask for an extended period. Theoretically, this condition will result in a gradual heat up of the cask. The thermal hydraulic analyses in Chapter 5 are carried out assuming that the duration of the STC-in-HI-TRAC condition is infinite so that steady state condition has been reached. Thus the "VCT breakdown" scenario is subsumed in the normal condition thermal analysis in Chapter 5. For site boundary dose calculations, the HI-TRAC cask is assumed to be between the two Fuel Buildings for 30 days (normal duration of transfer will be a few hours in one work shift).

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-9 Rev. 1

(i) Collapse of the roadway during transfer, resulting in a cask rollover [F.G]: This event is considered to be non-credible based on the following discussion.

First, the collapse of the roadway is considered to be a non-credible accident condition for the following reasons:

(1) The VCT will travel an approved route starting at the IP-3 cask loading area and traveling to the IP-2 cask loading area that will be evaluated (ground penetrating radar and/or soil compaction studies) prior to the transfer of the spent fuel. Since the site is situated on bedrock there is already an excellent base.

(2) The roadways at the site are qualified for H20 loads (typical tractor trailer loads) and the VCT loads on the roadway are similar, however the roadway will be upgraded to support the VCT with the loaded STC in the HI-TRAC. This upgrade includes placing concrete runways along the path for the VCT tracks to ride on.

(3) A portion of the load path has already been analyzed and used multiple times as part of the IP-2 and IP- 1 dry storage campaigns. That analysis is considered bounding since the VCT carrying a loaded MPC inside the HI-STORM weighs more than the VCT carrying a loaded STC inside the HI-TRAC.

(4) Prior to each transfer, the roadway will also be visually inspected and repaired as necessary.

Second, in the unlikely event that the roadway were to collapse, the VCT can withstand an eight foot depression of the roadway, in the most limiting configuration (one track of the VCT being eight feet above the other), without tipping over. This orientation of the VCT is considered bounding since the tracks of the VCT are longer than the width of the VCT. As can be seen in Figure 3.2.1 (a) the HI-TRAC is fully restrained in the VCT to stabilize the HI-TRAC during the transfer. Figure 3.2.1 (b) shows the VCT with eight feet of ground removed from under one track. Even in this most extreme scenario the center of gravity (c.g.) of the VCT carrying the HI-TRAC remains 1ow enough so that a tip-over (c.g. over corner) can not occur.

(j) Large radioactive release from the cask [F.G]: The STC is designed to hold a maximum of 12 assemblies during each fuel transfer. Chapter 7 provides the results of an analysis which hypothetically assumes all 12 of the assemblies release their gases into the STC cavity. The release fractions for the radioactive gasses are in accordance with the guidance in SFST ISG-5. This is more conservative than Regulatory Guide 1.183. No credit is taken for the seals' ability (STC or HI-TRAC) to retain the gases such that it is assumed all the gases are further released to the atmosphere. It is shown that the fuel handling accident presented in Chapter 14 of the UFSAR bounds this accident condition.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-10 Rev. 1

3.2.4 Load Cases Based on the design, normal condition, and accident condition loads identified in the preceding subsections, a series of governing load cases that require structural analysis are defined in Table 3.2.4.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-11 Rev. 1

Table 3.2.1 DESIGN PRESSURE AND TEMPERATURES Component Normal Pressure, Accident Pressure, Design Temperature, psig psig OF STC 50 65 Table 3.1.1 HI-TRAC 100D 30 40 Table 3.1.1 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-12 Rev. 1

Table 3.2.2 LIMITING OPERATION PARAMETERS Item Limit Maximum permissible lift height of HI- 6 inches.

TRAC 10OD in the VCT Maximum permissible plastic deformation 4.125 inches.

of the fuel box cell wall (downwards) from top of box cell Bounding weight of the fuel assembly 2000 lb.

Minimum ambient temperature 0 OF Maximum ambient temperature 100 OF Weather Forecast No snow, rain or lightning. Maximum wind speed < 20 mph Haul Path Connection No ice or snow on the haul path Design Basis Earthquake (ZPAs) Horizontal: 0.15g's I Vertical: 0.1Og's HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-13 Rev. 1

Table 3.2.3 DEFINITION OF THE FIRE CONDITION LOADING*

Fire duration 4.8 minutes Flame temperature 1475 0 F Maximum Ambient Temperature 100°F

Based on the fire event data in the HI-STORM 100 FSAR HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-14 Rev. -1

Table 3.2.4 GOVERNING CASES AND AFFECTED COMPONENTS Case Loading Event Affected Components Objective of the Analysis HI-TRAC STC VCT I Design Internal Pressure X X Demonstrate that the HI-HI-TRAC and STC under the TRAC and STC meets "ND" Design Internal Pressure stress limits.

2. Normal Operating Pressure plus X X - Demonstrate that the HI-Temperature TRAC and STC meets "ND" HI-TRAC and STC under stress limits.

normal operating pressure plus temperature 3 Normal Handling X X - Demonstrate that the Lifting of HI-TRAC and STC acceptance criteria in including dynamic effects Subsection 6.2.3 will be met.

4 Fuel Assembly Drop Accident -- X - Demonstrate that the A dropped fuel assembly plus acceptance criteria in handling tool impacts the top of Subsection 3.2.3.1 will be met.

the fuel basket.

5 HI-TRAC Vertical Drop X - -- Demonstrate that the peak Accident deceleration is less than 45g.

Vertical end drop of loaded HI-TRAC from maximum lift height

6 Seismic Stability of Loaded - - X Demonstrate that the loaded VCT VCT will remain stable under Loaded VCT subjected to DBE conditions.

Design Basis Earthquake (DBE) 7 Seismic Stability of Loaded HI- X - -- Demonstrate that the loaded TRAC HI-TRAC will remain stable Loaded HI-TRAC subjected to under DBE conditions.

Design Basis Earthquake (DBE) 8 Seismic Stability of Loaded STC -- X -- Demonstrate that the loaded in Fuel Pool STC will remain stable under Loaded STC subjected to Design DBE conditions.

Basis Earthquake (DBE)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-15 Rev. 1

Figure 3.2.1 (a) - HI-TRAC restrained in a VCT Figure 3.2.1 (b) - HI-TRAC on VCT with 8 foot roadway depression HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 3-16 Rev. 1

CHAPTER 4: CRITICALITY EVALUATION 4.1 Introduction This chapter documents the criticality safety evaluation for the Shielded Transfer Canister (STC) fuel basket containing either fresh or spent fuel assemblies with a nominal initial enrichment of up to 4.95 wt% 235U (+0.05 wt% manufacturing tolerance) in one of two analyzed storage arrangements:

Configuration 1 is analyzed to accommodate fuel with a specified minimum burnup as a function of the initial enrichment (see Table 4.7.3) in every basket location (see Figure 4.5.1)

Configuration 2 is analyzed to accommodate fresh fuel in the peripheral eight fuel basket locations (see Figure 4.5.2). The central four locations remain empty.

Each configuration is analyzed to demonstrate that keff is less than or equal to 0.95 with the storage fuel basket loaded with fuel of the highest anticipated reactivity and the STC flooded with water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95% probability at a 95% confidence level.

Under normal conditions, the water in the canister is assumed to be pure, unborated water, while under accident conditions, the soluble boron in the water is credited. A summary of the types of accidents analyzed and the soluble boron required to ensure that the maximum keff remains below 0.95 are shown in Table 4.7.4. These acceptance criteria are in accordance with 10CFR68(b).

Additionally, this chapter evaluates the acceptability of storing fuel from Indian Point Unit 3 in the pool of Indian Point Unit 2 (see Section 4.8).

4.2 Methodology The principal method for the criticality analysis of the high-density fuel basket is the use of the three-dimensional Monte Carlo code MCNP4a [V.B]. MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP4a was selected because it has been used previously and verified for criticality analyses and has all of the necessary features for this analysis. MCNP4a calculations used continuous energy cross-section data predominantly based on ENDF/B-V and ENDF/B-VI. PROPRI ETARY T EX REM'VOVED.,

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-1 Rev. 1

number of cycles and (4) the initial source distribution. The MCNP4a criticality output contains a great deal of useful information that may be used to determine the acceptability of the problem convergence. This information has been used in parametric studies to develop appropriate values for the aforementioned criticality parameters to be used in fuel basket criticality calculations. Based on these studies, 10,000 histories were simulated per cycle, 40 cycles were skipped before averaging, 160 cycles were accumulated, and the initial source was specified as uniform over the fueled regions (assemblies). Further, the output was reviewed to ensure that each calculation achieved acceptable convergence. These parameters represent an acceptable compromise between calculational precision and computational time.

CASMO-4 is used in this application to determine reactivity differences for moderator temperature variation, manufacturing tolerances, depletion uncertainty and to calculate the isotopic inventory of the spent fuel for use in MCNP4a. References [V.E] and [V.F] are Studsvik proprietary documents related to the appropriateness of CASMO-4 for calculating the multiplication factors.

Fuel depletion analyses during core operation were performed with CASMO-4 (using the 70-group cross-section N-library), a two-dimensional multigroup transport theory code based on the Method of Characteristics [V.D]. Detailed neutron energy spectra for each rod type are obtained in collision probability micro-group calculations for use in the condensation of the cross sections.

CASMO-4 is used to determine the isotopic composition of the spent fuel. In addition, the CASMO-4 calculations are restarted in the fuel basket geometry, yielding the two-dimensional infinite multiplication factor (kinf) for the fuel basket to determine the reactivity effect of fuel and basket tolerances, moderator temperature variation, and to perform various sensitivity studies. For all calculations, the Xe- 135 concentration in the fuel is conservatively set to zero.

Benchmark calculations for MCNP4a, presented in Appendix 4.A, indicate a bias of 0.0012 with an uncertainty of_+0.0090, are evaluated with a 95% probability at the 95% confidence level [V.A].The calculations for this analysis utilize the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix 4.A. Benchmark calculations for CASMO-4, presented in Appendix 4.B, indicate a PROPRIETARY TXT REMOVED bias and a bias uncertainty of PROPRIET ARY TEXT REMO , evaluated with a 95% probability at the 95%

confidence level [V.A]. Since CASMO-4 is only used to determine reactivity differences, the bias does not need to be applied to the results of the calculations. However, the bias uncertainty is included with the other uncertainties when determining the maximum kIff values.

The maximum kIff is determined from the MCNP4a calculated ker, the MCNP calculational bias, the temperature bias, and the applicable uncertainties and tolerances (MCNP and CASMO bias uncertainties, MCNP calculational uncertainty, basket tolerances, fuel tolerances, depletion uncertainty) using the following formula:

2 Max keff= Calculated kff + biases + [.i (Uncertainty) 21 4.3 Acceptance Criteria HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-2 Rev. 1

The objective of this evaluation is to show that the effective neutron multiplication factor, kff, is less than 0.95 with the fuel basket loaded with fuel of the highest anticipated reactivity and the STC flooded with unborated (normal conditions) or borated (accident conditions) water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95% probability at a 95% confidence level [V.A]. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95 under borated conditions. These acceptance criteria are in accordance with 10CFR68(b).

Applicable codes, standards, and regulations or pertinent sections thereof, include the following:

Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62, "Prevention of Criticality in Fuel Storage and Handling."

USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Rev. 3 - March 2007.

L. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T. Collins, August 19, 1998.

USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2, March 2007.

ANSI ANS-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements."

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-3 Rev. 1

4.4 Assumptions The criticality analyses use a range of assumptions, in order to simplify the calculations and/or to provide additional conservatism. In summary, those assumptions assure that the true reactivity will always be less than the calculated reactivity. The following is a list of the major assumptions, that were employed:

1. Moderator is water at a temperature that results in the highest reactivity, as determined by the analysis (see Section 4.7.7).

2. Neutron absorption in minor structural members is neglected; PROPRIETARY TEXT REMOVED~.

3. The fuel basket neutron absorber is 149 inches long, which is longer than the active region of the fuel of 144 inches. However, the absorber is conservatively modeled to be the same length as the active region of the fuel.

4. A conservative cooling time of 0 hours0 days 0 hours 0 weeks 0 months is used along with setting the xenon concentration to zero for all CASMO-4 calculations in the fuel basket models.- No credit is therefore taken for the significant cooling time of the fuel assemblies.

5. PROPR I4R/\Y TEXT RLNJOVED.

6. A bounding nominal fuel pellet density (96.5% theoretical) is conservatively considered in the analysis over the entire fuel rod length, i.e. fuel pellet dishing, chamfering or pellets with an annulus are conservatively modeled as solid fuel pellet cylinders with that density.

7. PROPRIETAkYTFXT REMOVFD.

8. PROPRIETARY TEXT 140 iD.

9. Reactivity control devices (RCCAs, WABAs, BPRAs, etc) that may be present in the fuel are not credited in the STC calculations.

10. A bumup record uncertainty of 5 % is assumed and incorporated in the analyses (See Table 4.7.3). For example, for an initial enrichment of 4.5 wt%, a fuel burnup of 35 GWd/mtU is required, but the analyses are performed at a burnup that is 5% lower, i.e. at 33.25 GWd/mtU.

11. All assemblies are assumed to be either centered in the basket cells, or all are assumed to be moved closest towards the center of the basket, whichever condition results in the higher klff value (see Section 4.7.5)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-4 Rev. 1

12. Calculations use nominal fuel and fuel assembly dimensions. Tolerances are treated as uncertainties (see Section 4.7.6) 4.5 Input Data 4.5.1 Design Basis Fuel Assembly Specification The STC fuel basket is designed to accommodate various Westinghouse designed 15x1 5 fuel assemblies used at Indian Point Unit 3. Data provided by Entergy encompasses various types of 15xI5 fuel assemblies (LOPAR, OFA, Vantage, Upgraded). The design specifications for these fuel assemblies are listed in Table 4.5.1. Specifications of the various inserts during core operation are summarized in Table 4.5.4 through Table 4.5.6.

4.5.2 Core Operating Parameters Core operating parameters are necessary for fuel depletion calculations performed with CASMO-4. The core parameters used for the depletion calculations are presented in Table 4.5.2.

The soluble boron concentration used bounds the cycle average for IP-3 for all cycles. The moderator temperature is the maximum (most conservative) over the core for IP-3. The neutron spectrum is hardened by each of these parameters, leading to a greater production of plutonium during depletion, which results in conservative reactivity results.

4.5.3 Axial Burnup Distribution PROPRI.ETARY andTR C SeMfi ctD 4.5.4 Fuel Basket and STC Specifications 11OPRIETARY TEXkT REMOVED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-5 Rev. 1

Table 4.5.1 Fuel Assembly Specification Assembly type 1 2 ' 3 15x15 Vantage 5, 15x15 15x15 Description Vantage +, Vantage LOPAR LOPAR P+iV+, Upgraded Fuel Fuel Rod Data Fuel pellet outside diameter, in. 0.3659 0.3649 Cladding inside diameter, in. 0.3734 Cladding outside diameter, in. 0.422 Cladding material Zr Stack density, g/cc (max) 96.5% TD Fuel Assembly Data Fuel rod array 15x15 Number of fuel rods 204 Fuel rod pitch, in. 0.563 Max. ZrB 2 Coating Loading (g 0.001043 (116 rods) or None

'°B/cm) 0.000871 (148 rods) None Max. ZrB 2 Coating Length, in. 128 None Number of Instrument/Guide 21 Tubes Guide Tube Material Zr Guide Tube inside diameter, in. 0.498 and0.499 0.512 Guide Tube outside diameter, 0.532 and 0.533 0.546 in.

Active fuel Length, in. 144 Axial Blankets Yes I No HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-6 Rev. 1

Table 4.5.2 IP-3 Core Operating Parameters Parameter Value Soluble Boron Concentration (cycle average), 900 ppm Reactor Specific Power, MW/MTU 36.80 (through Cycle 15)

Core Average Fuel Temperature, 'F 1280 Core Average Moderator Temperature at the 620 Top of the Active Region, 'F In-Core Assembly Pitch, Inches 8.466 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-7 Rev. 1

Table 4.5.3 Axial Burnup Distribution [V.1]

IPRd1RIEFTA R ~ P'OPRET!AR) TE-XT REMOVI)\EP TEXT ~REFMO10VFh P~ROPRIETARY TEXT PkdIPRIETARYV PROPRIE AY PROPRIETARY REFNFRMOVED) MOMED TEXT TEIT(EMVE Q TEXT, REMOVE 1 PROP13IETARY~TEX PROPR[ETARY PROPR1ETAI \R PROPRIETARY REMOVED J IXTF REMOVE)ID. IEX I REMOVD TEXT REMO~VEI PROPRIETA R TEfvl PROPRWETARV PR OPH IET AR1ý) PROPRIETARY REOVDTEXT REMOVED, TEXT REMOVMED, TEXT. REMOVMED PRO PRIETARKYTEXTI PROPR1ETýRY PROPR~IETA`RY 'PROPRIETARY-kEMOVED' TENXT4F(EMDQYE RL%0\ID TEXT REMOVEDTE PRO~PRIETARY ItEXT PROPR LTA10' PROIETJARY P>ROPRIETARY~

REMOVED TEXT R~EMOVED TEXT REM~OVED TEXT REMOVED PROPR~IETARY ITEXN PROPRITAR PROPRIETARY- PROPRIETARY

-REMOVED, TEXI REM%()\ ED 1RE O/ED.

ITEXT 'TEXT REMVOVED PRORE~TARY TEXT PROPRIETARY~ PROPRkIETARY- PROPRIE~TAR.Y R-EMOVED jTEXTrRE'MOVEQ TEXT REMOVED TEXT REMOVED TEXP~rAYT POPRIETARY, PROPRETAR~Y- PR7IOPRIETARY, Riý\MOVED, f EXT[REMOI()VE TEXT R-EMOVED, TEXT REMOVED PROPRIETARY TET PP(R f FTARý PROPRIETARYj PROPRIETAR' REMkOVED. TEXT REMOVED TE~XT REM~OVED~ TEXT R X-lOED PROPRIET~AkY TEXT PROPPIFT7ARY PRtOPRIETARY, PROPRIETARY(

_R.E*4QVE -TE,\TREMOYE 1) TEXT~REMOVED I XI REMOVEDJ POPRtETAý TPIX7 1 PROPR~IETARY, PROPRIETARY PROiPRIETARY~

REMOVED, TEXT REMOVED~ TEXT REMOVED, TEXT REMOVED PROPRIETARY' fXT PROPRIFTRY P~ROPIiETARY PROPPIETA1RY REM\OVEI) -1EXT REMOV()\ED T~EXT ~REMOvED E)XT\

REMOVED PROPRIETARY, TEXT 1RO~PRIE ,AkýR, PROPRI1ETAPY PROPRIETARY REMOVED{ TEXT REMOVED TEXT REM~O~ED TEXT. REMOVED PROPRIETARY TEXT PRO~PR IETARY I EROPRIETARY" IPOPRIETý\k REMOVED TEXTIRMVED TEXT REMOVED TlEXT RA(ATE PROPRIETA RY"f1 TEX PROPRIETARY PRPITR PROPRIL FARY REMOVEI) TEX~T REMOVED TEXT REMOVED TEXTREMQVED PRPREAPý AT POPIEAY RORETR PROPRIETARY PROPRIETARY TEXT~ PROPRIETARY' POREAY RPITR

-REMOVED TEXT REMOVED jTEXTI'MOVED TEXT REMOVED PROPRI ETARY TEXT1 PROPRIETARY PROPRIETARY PROPRIETAR'Y REMOVED TEXT REMOVED TEXT RE MOVED TEXT REMOVED PROPRIETARW"IFTE PROPRIETARY PROPRIE~TARY, PiOPR41 FYý

-;EMOVEDI jTEXTREM\OVEb TEXT REMOVED I EXT REMOVED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-8 Rev. 1

Table 4.5.4 Burnable Poison Rods Parameter Value (Cycle 1-4) Value (Cycle 8-10)

Max Number of BPRs per 20 20 assembly BPR Inner Clad ID (in.) 0.2235t 0.2230 BPR Inner Clad OD (in.) 0.2365 0.2360 BP ID (in.) 0.2450 0.2440 BP OD (in.) 0.3920 0.3890 BP Outer Clad ID (in.) 0.4005 0.3935 BP Outer Clad OD (in.) 0.4390 0.4310 Cladding Material Stainless Steel Stainless Steel Poison Material (Borosilicate B 2 0 3 -SIO 2 (18.1 wt% B 2 0 3 ) B 20 3-SIO 2 (12.5 wt%

Glass) Cycle 1-2 B 2 0 3)

B 2 0 3 -SiO 2 (12.5 wt% B 20 3)

Cycle 3-4 Burnable Poison Density 2.23 2.23 3

(g/cm )

AssemblyBurnup when 30 19 absorber is removed (GWD/MTU)

ý Analyses used 0.2350 Inch. This has a negligible effect.

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Table 4.5.5 Wet Annular Burnable Absorbers (WABAs)

Parameter Value Max Number of rodlets per assembly 20 WABA Inner Clad ID (in.) 0.2210 (Cycle 5-6) 0.2250 (Cycle 7-15)

WABA Inner Clad OD (in.) 0.2670 A120 3-B 4C ID (in.) 0.2780 A12 0 3 -B 4 C OD (in.) 0.3180 WABA Outer Clad ID (in.) 0.3290 WABA Outer Clad OD (in.) 0.3810 Cladding Material Zr Poison Material 0.00603g l°B/cm Max Absorber Length (in.) 134 Assembly Burnup when Absorber 33.1 is removed (GWD/MTU)

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Table 4.5.6 Hafnium Flux Suppressor Rods Parameter Value Max Number of rodlets per assembly 16 Hafnium Rod OD (in.) 0.3810 Poison Material Hf Absorber Length (in.) 143 Assembly Burnup while absorber is 6.1 present (GWD/MTU)

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Table 4.5.7 STC and Fuel Basket Specification Parameter Value Cell ID, Inches PROPRIETARY TEXT Box Wall Thickness, Inches P1RORIE YARYTEX REMOVE12 Cell Pitch, Inches PROPIETA Sheathing Thickness, Inches 1RRIETAR TA'F REMO\,F Metamic Poison Thickness, Inches PROPRIETARY TEX R~EMOVED Metamic Poison Width, Inches PROPRTARY TE REMOVED1 Metamic Poison B 4 C Weight Percent PROETA TEXT

________________________, REMOVED~

PROPRIETR TFX STC ID, Inches ........... Th),

STD Wall Thickness (inside to outside), Inches PRdPL T XI

" 'ARYTI-RV-MVED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-12 Rev. 1

Figure 4.5.1: A Two-Dimensional Representation of the Actual Calculational Model Used for the Configuration 1 STC Analysis for Spent Fuel (To Scale).

(FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

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Figure 4.5.2: A Two-Dimensional Representation of the Actual Calculational Model Used for the Configuration 2 STC Analysis for Fresh Fuel (To Scale).

(FIGURE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

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4.6 Computer Codes The following computer codes were used during this analysis:

MCNP4a [V.B] is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory. This code offers the capability of performing full three-dimensional calculations for the loaded Fuel Basket. MCNP4a was run on the PCs at Holtec.

Continuous energy cross-section data was based on ENDF/B-V and ENDF/B-VI.

CASMO-4, Version 2.05.14 [V.D] is a two-dimensional multigroup transport theory code developed by Studsvik of Sweden. CASMO-4 performs cell criticality calculations and burnup. CASMO-4 has the capability of analytically restarting burned fuel assemblies in the fuel basket configuration. This code was used to determine the reactivity effects of tolerances and fuel depletion and to determine the isotopic inventory of spent fuel. The N-library was used for all calculations.

4.7 Analysis Full three-dimensional calculational models were used in MCNP, explicitly modeling fuel rods and cladding, guide tubes, basket walls, and neutron absorber panels on the basket walls covered by sheathing. During the stages of the transfer operation, the STC can be in one of four configurations: 1) Inside Unit 3 pool; 2) Inside Unit 2 pool; 3) Elevated by a crane hanging in air; and 4) inside the water-filled HI-TRAC. To appropriately represent and bound those configurations, the STC around the basket is included in the model, surrounded by a water reflector of more than 12 inches on the side, top and bottom, but the HI-TRAC is not included.

This is appropriate, since the 12 inches of water essentially represent full reflection on the outside of the STC, and any further reflection from the water of the Unit 2 or Unit 3 pool, or from the HI-TRAC body would be negligible. Further, it also bounds the condition of the STC in air, which would result in less reflection. Section 4.7.10.1 present studies that confirm this approach is appropriate. Figures 4.5.1 and 4.5.2 show cross sections of the models for Configuration 1 and 2, respectively. CASMO uses infinite arrays of basket cells in a two-dimensional geometry.

Unless otherwise stated, all calculations assumed nominal dimensions for the fuel and the fuel storage cells. The effect of the manufacturing tolerances is accounted for within the uncertainty analysis as discussed below.

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4.7.1 Fuel Assemblies, Bumable Poison, and Hafnium Inserts Effect of Assembly Types The reactivity of the three assembly types, with and without inserts, has been evaluated and the results are listed in Table 4.7.7. In all cases, assembly type 1 results in the highest kif value. This assembly type is therefore used in all further calculations.

Effect of Hafnium Inserts P'ROP~RIETARY TEXTRFMOV& D 4.7.2 Reactivity Effect of Axial Bumup Distribution Initially, fuel loaded into the reactor will burn with a slightly skewed cosine power distribution.

As burnup progresses, the relative bumup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. Note that the relative distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced bumup. In any case, the more reactive fuel near the ends of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage. Previous analyses have generally shown a negative reactivity effect of the axially distributed burnup compared to a flat distribution at low bumups, becoming typically positive at bumups greater than about 20-25 GWD/MTU. The required burnup for the maximum enrichment is higher than 25 GWD/MTU. Therefore, a positive reactivity effect of the axially distributed bumup is possible. PROPRIETARY TEXT REM@1(VEID' 4.7.3 Isotopic Compositions Fko F~RI 7T'R\ 1 tEXT' REMO ED',

4.7.4 Uncertainty in Depletion Calculations 4.P7 IEcAce TEXT Ful Tsem.

4.7.5 Eccentric Fuel Assembly Positioning HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-16 Rev. 1

Normally, the fuel assemblies are expected to be located at the center of the fuel basket cell. To investigate the potential reactivity effect of eccentric positioning of assemblies in the cells, additional MCNP4a studies were performed. For eccentric positioning all the fuel assemblies are positioned toward the center of the fuel basket The results of these studies are presented in Table 4.7.10 and indicate that in most cases the eccentric fuel positioning with the assemblies placed closest to the center of the basket result in an increase in reactivity. In any case, the condition with the highest reactivity shown in that table is used in the design basis calculations.

4.7.6 Uncertainties from Manufacturing Tolerances In the calculation of the final kff, the effect of manufacturing tolerances on reactivity is included as discussed in Section 4.2. CASMO-4 was used to perform these calculations. As allowed in

[V.G], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. The evaluations include tolerances of the fuel basket dimensions and tolerances of the fuel dimensions.

The calculations are performed for different enrichments (1.8 to 4.95 wt% 235U) at various burnups and with a soluble boron concentration of 0 ppm and 600 ppm. The license limit of 5.0 wt% 235U is the sum of the nominal enrichment and the manufacturing tolerance. Therefore, the maximum nominal enrichment of 4.95 Wt% 2 35U allows for a manufacturing tolerance of 0.05 wt% 235U. To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kinf from a calculation with the tolerance included. Note that for the individual parameters associated with a tolerance, no statistical approach is utilized. Instead, the full tolerance value is utilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for fuel assembly and fuel basket manufacturing tolerances. A maximum fuel and basket tolerance is determined over all relevant burnup and enrichment combinations, and is used in all design basis analyses. The fuel and basket tolerances included in this analysis are described below. The fuel tolerances are assumed values typically used for those dimensions; the cell ID and cell pitch tolerances are taken from the drawings in Chapter 1; the box wall and sheathing thickness tolerances are based on the tolerances for steel sheet material; and the Metamic thickness tolerance is from the manufacturing specification of the material. Note that for Metamic, the design basis calculations already use the minimum B 4 C content and minimum panel width, therefore, only the effect of the thickness tolerance is evaluated.

Fuel Tolerances 3

Increased Fuel Density - + 0.1096 g/cm

Increased Fuel Enrichment - + 0.05 wt% 2 3 5 U

" Fuel Rod Pitch - +/- 0.001 in.

o Fuel Rod Cladding Outside Diameter - +/- 0.00 15 in.

" Fuel Rod Cladding Inner Diameter - +/- 0.0015 in.

. Fuel Pellet Outside Diameter- +/--0.0005 in.

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" Guide Tube Outside Diameter - +/- 0.003 in.

Guide Tube Inside Diameter - +/- 0.003 in.

Fuel Basket Tolerances

Cell Inner Dimension& Pitch - PR&PRIETARYTEXT 1[MOVED.

Box WallThickness - PRQPRE2ARYtIFXTREMOVED.

" Sheathing Thickness - PRO ETARY TEXT REM ED1

Metamic Thickness - PROPRIETARYTEX MOED.

Only the Ak values in the positive direction (increasing reactivity) were used in the statistical combination. The results are summarized in Table 4.7.9. The table shows the individual and combined uncertainty for fresh and maximum burned fuel at 4.95 wt% enrichment for 0 ppm soluble boron, and for the enrichment and burnup combinations that result in the highest combined values for 0 and for 600 ppm soluble boron.

During manufacturing there is a potential for minor damage to the neutron absorber panels from welding the sheathing to the cell walls. Such damage, up to an area equivalent to 1 inch diameter per panel, was considered in Holtec's HI-STAR Transport SAR [K.C], Section 6.4.11, for various baskets similar to the STC basket, and was found to be acceptable. This condition is therefore also acceptable for the STC, without any further calculations.

4.7.7 Temperature and Water Density Effects Water temperature and density effects on reactivity in the STC fuel basket have been calculated with CASMO-4 for various enrichments with a maximum value of up to 4.95 wt% 235U and the results are presented in Table 4.7.8. The results show that the spent fuel water temperature coefficient of reactivity is negative, i.e., a higher temperature results in a lower reactivity, and that a reduced density results in a reduction in reactivity. Consequently, all CASMO-4 calculations are evaluated at 39.2 'F (4 °C), and full density water.

In MCNP4a, the Doppler treatment and cross-sections may only be valid at 300K (80.33 'F).

Therefore, a Ak is determined in CASMO-4 from 39.2 'F to 80.33 'F, and is included in the final kff calculation as a bias.

As for the uncertainties, a temperature bias is determined for 0 ppm and 600 ppm soluble boron.

During the entire transfer operation, the active fuel region remains covered with water. However, should there be any condition in which part of the active fuel is not covered by water, then that will not have any adverse effect from a criticality perspective, since the calculations show that a reduction in density results in a reduction in reactivity. This is consistent with previous studies that show that for baskets with neutron absorbers, the optimum moderation condition corresponds to full water density.

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4.7.8 Calculation of Maximum kff for Normal Conditions Using the calculational model shown in Figure 4.5.1 and 4.5.2 and the design basis fuel assembly specified in Table 4.5.1, the klff in the fuel basket has been calculated with MCNP4a for both configurations. The determination of the maximum keff values, based on the formula in Section 4.2, was calculated for an initial enrichment of 4.95 wt% 235U for Configuration 2 and for initial enrichments between 1.8 wt% 235U and 4.95 wt% 235U, and the corresponding burnup listed in Table 4.7.3, for Configuration 1. The results show that the maximum kff of the fuel basket loaded in accordance with either Configuration 1 or Configuration 2 is less than 0.95 at a 95%

probability and at a 95% confidence level without credit for soluble boron. See Table 4.7.1 and Table 4.7.2 for detailed results.

4.7.9 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. This section identifies which of the credible abnormal or accident conditions will result in exceeding the limiting reactivity (keff < 0.95). For those accident or abnormal conditions that result in exceeding the limiting reactivity, a minimum soluble boron concentration is determined to ensure' that kff < 0.95.' The double contingency principal of ANS-8. l/N16.1-1975 [V.H]

specifies that it shall require at least two unlikely, independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions. For those cases where the reactivity of the accident is expected to be greater than the limit of kff *- 0.95 with no soluble boron credit, calculations were performed with soluble boron and the concentration required to meet the limit is specified.

4.7.9.1 Abnormal Temperature All calculations for the fuel basket are performed at a water temperature of 39.2 'F (4 'C). As shown in Section 4.7.7 above, the temperature coefficient of reactivity is negative; therefore no additional calculations are required, because a further increase in temperature reduces the reactivity.

4.7.9.2 Misplaced Assembly Dropped Assembly For the case in which a fuel assembly is assumed to be dropped on top of the STC, the fuel assembly is assumed to come to rest horizontally on top of the STC, with a minimum separation distance from the active fuel region of more than 12 inches, which is sufficient to preclude neutron coupling (i.e.,

an effectively infinite separation). Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. It is also possible to vertically drop an assembly into an empty location or a location occupied by another assembly. Another condition that could potentially HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT Hl-2094289 4-19 Rev. 1

result in minor damage to fuel assemblies would be the drop of the HI-TRAC during lifting operations with the VCT. Such vertical impacts would at most cause a small compression of the assembly, reducing the water-to-fuel ratio and thereby reducing reactivity. Furthermore, the reactivity effect of a dropped assembly would always be bounded by the misloading condition discussed below, and the soluble boron maintained in the spent fuel pool water in accordance with the plant technical specifications assures that the true reactivity is always less than the limiting value for such dropped fuel accident.

Mislocated Assembly The spaces between the basket and the inner diameter of the STC are too small for a fuel assembly. Mislocation of an assembly on the outside of the basket is therefore not credible.

Misloaded Assembly A misloading condition is analyzed for both loading configurations of the basket. For both conditions, it is assumed that a fresh assembly with an enrichment of 4.95 wt% is loaded in one of the four center cells of the basket. For Configuration 1, this assembly replaced a spent fuel assembly, while for Configuration 2, it is placed into one of the empty cells at the center of the basket. Without credit for the presence of the soluble boron in the water, the maximum keff exceeds the limit of 0.95. Additional calculations were therefore performed which credit the presence of soluble boron in the water. Results for those misloading conditions are summarized in Table 4.7.11 (Configuration 1) and 4.7.12 (Configuration 2). With the soluble boron levels listed in Table 4.7.4, the maximum keff is below the limit of 0.95.

4.7.10 Additional Sensitivity Studies 4.7.10.1 External Reflection The design basis model includes the STC surrounded by more than 12 inches of water on all sides. To show that this bounds all expected actual configurations (see Section 4.7), a study is performed with variations in the external conditions. The configurations analyzed are

specifying a reflective boundary condition at the outer surfaces of the water reflector, thus increasing the neutron reflection; and

replacing the water reflector with a void, thus reducing the neutron reflection.

Results are listed in Table 4.7.13. They show that the differences in external reflection have a negligible effect on the kff of the system, and that the design basis calculations are bounding.

The design basis model is therefore appropriate for all expected conditions of the STC during transfer.

4.7.10.2 Margin to Account for Additional Uncertainties that are Difficult to Quantify The methodology, specifically for evaluating spent fuel, may have additional uncertainties that are difficult to quantify. These may include uncertainties in the depletion analyses beyond the 5% currently considered. However, the calculations also apply numerous conservatisms to offset HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-20 Rev. 1

such additional uncertainties. To show the effect of those conservatisms, an additional calculation was performed where some of those conservatisms were reduced. Specifically, the following changes were made:

Depletion calculations were performed for a cooling time of 5 years, which is the minimum cooling time required for the fuel in the STC. The design basis calculations assume no cooling time.

" Moderator temperature during depletion is set at 605 F, which is still above the typical core outlet temperature, but below the value of 620 F used in the design basis calculations.

For the Pyrex inserts during depletion, the design basis calculations assume a B20 3 content of 18.1 wt%. This value was only used in the very first cycles, i.e. for assemblies that by now have a substantial cooling time. Later designs used a value of 12.5 wt%. This value is used in the evaluations presented here.

The result of this calculation is presented in Table 4.7.14, in comparison with the corresponding design basis calculation. The comparison shows that the reduced conservatisms result in a reduction of the keff value of 0.0157. This is comparable to the uncertainties already considered (see Table 4.7.1), and therefore would ensure that any additional uncertainties not explicitly evaluated are covered by the evaluations.

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Table 4.7.1 Summary of the Criticality Safety Analyses for Configuration 1, Normal Condition Design Basis Burnup at 4.95 Wt% 235U 38.0 GWD/MTU Soluble Boron 0 ppm Uncertainties MCNP Bias Uncertainty (95%/95%) +/- 0.0090 CASMO Bias Uncertainty (95%/95%) IPROPRIFT,ý\kY TEXT REM 'lED Calculational Statistics (95%/95%, 2.Oxcy) +/- 0.0014 Fuel Eccentricity Included Basket Tolerances +/- 0.0036 Fuel Tolerances + 0.0087 Depletion Uncertainty +/- 0.0106 Statistical Combination of Uncertainties +/- 0.0172 Reference kf (MCNP4a) 0.9118 Total Uncertainty (above) 0.0172 Temperature Bias 0.0048 IFBA Bias PROPIAI0.00 12 Calculational Bias (see Appendix A) 0.0012 Maximum keff 0.9379 Regulatory Limiting keff 0.95 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-22 Rev. 1

Table 4.7.2 Summary of the Criticality Safety Analyses for Configuration 2, Normal Condition Design Basis Burnup at 4.95 wt% 235U 0.0 GWD/MTU Soluble Boron 0 ppm Uncertainties MCNP Bias Uncertainty (95%/95%) + 0.0090 CASMO Bias Uncertainty (95%/95%) PROPRIETAf) ~TEX I-REMO V1f' Calculational Statistics (95%/95%, 2.Oxcy) +/- 0.0014 Fuel Eccentricity Included Basket Tolerances + 0.0036 Fuel Tolerances + 0.0087 Depletion Uncertainty N/A Statistical Combination of Uncertainties +/- 0.0136 Reference kff (MCNP4a) 0.9127 Total Uncertainty (above) 0.0136 Temperature Bias 0.0048 Calculational Bias (see Appendix A) 0.0012 Maximum keff 0.9323 Regulatory Limiting keff 0.95 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT 1-11-2094289 4-23 Rev. 1

Table 4.7.3 Burnup Versus Enrichment Requirement for Configuration 1 Burnup (G D/MITU)

Limit for Enrichment Used in Assembly (wt% 23 5U) Analyses Selectiont 1.8 0 0 2.0 5.22 5.50 2.5 12.11 12.75 3.0 17.29 18.20 3.5 22.70 23.90 4.0 28.26 29.75 4.5 33.25 35.00 4.95 38.00 40.00 t a) For fuel assemblies irradiated with a Hafnium insert, the assembly burnup prior to the Hafnium insertion is to be used for comparison with the values in this column.

b) Minimum bumup values at intermediate enrichments can be determined by linear interpolation HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-24 Rev. 1

Table 4.7.4 Summary of Accident Conditions for the STC Soluble Boron Requirement, Case ppm Abnormal Temperature Negative Dropped Assembly - Vertical Negligible Dropped Assembly - Horizontal Negligible Misloaded Fresh Fuel Assembly in Configuration 1 600 ppm Misloaded Fresh Fuel Assembly in Configuration 2 600 ppm HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-25 Rev. 1

Table 4.7.5 Reactivity Effect of Insert Types HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-26 Rev. 1

Table 4.7.6 Reactivity Effect of IFBA Rods IPROPRIETARY PPOPRITAR~ PROPRIETAR Y TEXT REMOV ED,____

IEOV FOE gXoR1ýF PROPRI EA ?40iý ROPRIEJ PROPRIET PROPR1E' REMORY TEIIY~ AYT[ x F TXTX J ARX ARY TEX FRTX REMOVED REMOVED REMOVED REMOVMED REMOVE P)ý1RO EAVT IPRIETARY PROPRIET.ARý PROPRIE~IkdkET~ PR P ',POREPORE'

ý1Ex i ITEX __ RY TEXT- ARY'TE% ,A ITFI ARY TEXT ARY TEXT

'REM 6OVEDiL REMOV IED REMOVE D REMOVE\ D;REMOVED I REMOVED REMOVEL)

PROPRIETAR) PROPRIETAR PROPRI4ETAI PROPRIE~ PIROPRIET, PROPRIE PROPRIETt

~TEXT ~ X'TEX1T R)xfK ARY'II K ARY TEXTl ARY' LX ARY TEXT~

REMOVED~ REMOVED MEIUEt REO§ ED, REMOVED REMOVED REMOVED, P~ROPRIETARY POPRIE~TAER PROPRIETA PROPRJIJT PROPRJETj PROPRIET PROPRIET, TrEXTI RMOVD HTEX RY TENT IRYxTu Y TEXT AIY

\ý IEx~ ARY TEXT4 IREMOVE,D, REMOVED, REMOVED~ REMOVED! REMOVED REMOVEDi PROPRIETARY PROPRIETARA PROPRIET.A ~PIROPIFT[ fROPRIET, PRO1PkIE~ iROPRIE~T

'I EXT jyIT-x1111 RýYTEXT AR), IEX'I ARý'EXT A R),TEXT ARY TEXT REM\Of1) P\0VED REMOVED REMYDREMOV~ED REMOVED, REMOVED REMOVED PROPRIETARY PRP 1E~FPROPRIETA~ PROPRIET PROPRIET IPROPRIET PROPRIET,

~TEXT Y XThX1j-11 R 1 LXTj ARAY TEXT ARY TEXT ' RY TEXT ARY TEXT REMOVED~ REMOVED, REMOVED, RM%11 REMOVED REMVED ' OVQED REMOVED PROPRIETARY~ PRO PIETAR~ PROPRIETA P1ROPRIE FPROPRIET, PROPRIE, PROPRIET~

TEXT ~ Y IIEX RY TEXT ARY TEXT ARY TEXTj ARY TEXT AkXVTEXT REMOVE[D REMOVELD REMOVED, REMOVED, REOE REMOVED, REMOVED PROPR~IETARY PROPRIETA1 PROPRIETAI PROPRfIET PROP.)I..I PRPR PROPRWET F[ A11111 TX{ RY` TEXT ARYTEX1T ARY TEX I ARY TEXT ARY TEXT REMOVD KhOVED ý_mý[I REM 6ED REMOVED, REMOVED OV~ REMOVED PRORIEAR __PITRPORI.LPORE PROPRIET PROPRIET~ PROPRIETj

,,TEXTYI xf iTE'PX3T III TEXý'EXAYIIýxI RY-¶EXTLI AYTET REM~OVED RýEMOVEDj REMOVED REMOVED REMOVE R)PEMOVED REOE PROPR1ETARýý PROPRIETAR~ QPROPET21A PROIPRIIET PROPRIET PR~OPRtET, PROPRIET REMOVED REMOVED REMOVED~ REMOVED R1-NMIOVE 1 REMOVED REMOVE 1 PR(WRIETARY~ PROPRIETAR PROPRýIETA PROPRJET PROPRIE FPROPRIET PROPRIET F 1TEXT IE~f RYTEIXfX ARY TEXT ARY TEX Aý

.RY TEXT ARY TEXTI R2 AKEMOE REMOVED RI NUNED REMOVED, REMOVED, REMOVED, REMOVED PIMRPRIETARY~PROPRIETAR ROPRIE<Al PROP~RIET PROPR1ET, PROPRIE FPR~OPRIET T7EXT1 TE~XTI RY TEXT1 ARY'TEXTj ARY TEXT YýTEYI ARY TEXT R~EM~OVED R~EMOVED, REMOVED~ REMNOVEDi REMOVED TREMVOVED, REMOVED.

PROPRIE3TARY 1PkOPRIETAR~ PROPRIETA PIROPRIET PORE ORIT RPIT T EXT, LX~ Y TEXTI AR)-TEXT, ARY' IEX N A RYTEXT, ARYTII ý REMOVED~ REMOVED IREMOVED IREMOVEDi REMOVMED -REMOVEQ, REMOVED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-27 Rev. 1

Table 4.7.7 Reactivity Effect of Fuel Types PfROPRIETARY PROPRIETARY '[EXT REMOVED PROPRIETAR PROPRIETARY PROPRIETARY

-TEX I TEXT TEXT REMOVED~ REMOVED~ REMOVE PROPRIETARY~ PROPRIETAR PR(IRITARY TEXT IETEI2FX REMOVED REMOVED REMýNOVED PROPRIETARY PROPRIETA4RY PROPRIETARY TEXT TEXT TEXT~

REMOVEPD REMOVED PROPRIETARY~ PROPRIETARY PROPRIETARY TEXF\tE XT

.REMOVED REMOVED REMhtOVED PROPRITR PROPRIETARY PROPRIETARY TEXT TE~XT TEXT REMIOVEDJ REMOVED REMOVED PROPRIL.1R PROPRIETARY PROP~RIETARY TEXT TEXT REMO\TE1 REMOVED REMOVE~D PROPRIETARY P~ROPRIETARY PRO~PRIETARY TEXT TEI f- TFX~T REFMOVED PEMOVEDi REM OVED PROPRIETARY PROPRIETARY PROPI~RETAR~Y TEXT TEXT REMOVED REMOVED PROPRIETARY PRO5PRJETARY PRO~PRIETARY TEXT TEX-ýX REMOVED R~EMOVED PROPRIETARY PR~OPRIETARY TEXT TEX TEXT

-REMOVED R-EMOVED REMOVED PROPRIETARY PROPRIETARY~ PROPRIETARY TEXT1 TEXT TEXT

-REMOVED REMOVED R-M~OVED PROPRIETARY PRO~PRIETARY PROPRIETARY TEXT I 1-ýTTEXT~

REMOVED RýE,%,VED~ REMOVEDi PROPRIETARY PROPRIETARY PROPRIETWARY TEXT TrEXT TEX REMOVEDi REMOVED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-28 Rev. 1

Table 4.7.8 Reactivity Effect of Temperature Variation in the STC Fuel Basket Temperature (OF) 4.95 wt% 1.8 wt% 1.8 wt% 4.95 wt%

0 GWD/MTU 0 GWDIMTU 0 GWD/MTU 40.0 0 ppm (Bounding, (Bounding, GWD/MTU 0 ppm) 600 ppm) 0 ppm Ak Ak Ak Ak 39.2 (4 °C) Reference Reference Reference Reference 68 (20 °C) -0.0020 -0.0033 -0.0023 -0.0015 80.33 (300K) -0.0031 -0.0048 -0.0033 -0.0023 150 (65 °C) (max -0.0104 -0.0143 -0.0092 -0.0081 normal temp) 254 (120 °C) -0.0252 -0.0310 -0.0191 -0.0197 254 + 10% Void -0.0487 -0.0495 -0.0267 -0.0420 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-29 Rev. 1

Table 4.7.9 Reactivity Effect of Fuel and Basket Tolerances for the STC Fuel Basket Tolerance 4.95 wt% 1.8 wt% 1.8 wt% 4.95 wt%

0 GWD/MTU 0 GWD/MTU 0 GWD/MTU 40 GWD/MTU 0 ppm 0 ppm 600 ppm 0 ppm Fuel Tolerance Ak Ak Ak Ak PýROPRWEARY TEXX PROP RIETk\ R P4RIETAR. ROPI'ETAR~ PROPRIETARY RFMOVED 'IEýXTfREMIOV L ý IE Y TEXTL TEX XT RFMOV~E 1 REMOVED~ REMOVED PR(WORI F'[ý' (JEX)l'l PILt'2Rý- PR ()I'lRI ETA PROPRIETAPI PROPRIIETARY~

IMOF.E EXTI REMOVEDi Ir~

LNI YTEX'1I TEXT REgIOXE7D R~(\QREMOVED EMOV, PROETPIMEt RPRIETA~RY PIMPRIETAk! PROPREk ETMPIARY REMOVED I,EX7K REMOVEb Y-TIXF Y TEXT TEX'T REMOVED kE'Mv 1( ED REMOVED PROPRIETARY4EE 'PRO(WRETARYL PR~OPRIETAR PROPRIET-AkP PROPRIETARY REMOVE~' 1)E1XTI RI \M) 1-) 'EXT_ Y TEXTF TEXT REMOVE REMOVED REMOVED, PROPRIETAR TFTEX, NW~i)RITARPY PROPRIETAR~ PROPRIETAPI PROPRIETARY REMOVE" TE1XTI REMIOVE Y))TEXT Y4'ET I ~ TEXT REMOVED REMOIVED REMOVED~

1PkOPRJETARýJE PIý PR.IIE] ARV PROPRIETAPI PROPRIETAJR PROPRIETARY REMOIVEP ILnX IRENIOYE D ' t EXTI Y F EX~T REMOVE~

___________ _RtAMOVED RM-\OVED

'NEPRIETARYEýI PROPaJETAP PROPRIEIý PROPRIETAR PROPRIETARY REMOVED TEf-XT REMO0VE Y T'EXT TEXVI~ TEXT REMOVED REMOVED b EMED PROPRIETARY TEX3 PROPRIIETAR RORAk PRPREAR PORIETARY J? %(),F) I x]REMOV'L ý'TEX T XL TEXTREOD REMOVED REMOVED Statistical Combination 0.0032 0.0087 0.0091 0.0034 4.95 wt%

0 GWD/MTU 600 ppm Basket Tolerance Ak Ak ' Ak Ak P RO P RIF kf, fýEXT PROPRIETARY PROPRIETA PROPRIETAR POPRITARP)

R FNIVOj

()\' ) N REMOVED~ Y TEXT, Y TEXIT TEX I EMOVAED~

REMOVE MO PR~OPRIETARt'Y'T'ENT, PR.OPRUIEARY PROPR iIETAR~ PIROPRIETAIR PROPRIETARY~

REM.OVED, [rEXJ REOE TEXT Y TEXT TEXT REMOVED REMOVEDi REMOVED PROPRIETARY TEXT] PROPRIETARY\ PROPRIETA1R PROPRIETAR JPROPRRETAR$'

REMOVED, TEXT REM"OVEDi Y El"XT Y TEXT TEXT REMOVED1

________________________REMOVED R~EMOVED

,!*OP~iETARYTEXT PROPRIETARY PROPRIETA PRPIEA PROIPRJETARY REMOVEf)[TEXTI 'R[EMOVEIJ ý -! YTEXT IYEX 1tXY REMOVED IREMOYED REMOVE Statistical Combination 0.0036 0.0020 0.0039 0.0026 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-30 Rev. 1

Table 4.7.10 Reactivity Effect of Eccentric Positioning Case Calculated keff Assemblies Moved to Assemblies Centered basket Center in Cells PROPRIETARY PROPRIET.UhY TEXT REMOVELi TEXT REMOVED~

PROPRIETAR PROPRIETARY TEXT R;FO,1Eii"F TEXT R-EMOVEDI XPROPRIFTAR PROQPRWIEARY TET PJMOVEED TEXT REMOVED PROPRIETARY PROPRIFTARY~

TEXT REMOVED~ TEXT~ RMOVED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-31 Rev. 1

Table 4.7.11 Summary of the Criticality Safety Analyses for Configuration 1, Accident Condition Design Basis Burnup at 4.95 wt% 23U 38.0 GWD/MTU Soluble Boron 600 ppm Uncertainties MCNP Bias Uncertainty (95%/95%) +/- 0.0090 CASMO Bias Uncertainty (95%/95%) kRfRMF1ET -TEXT, R-EMOVED Calculational Statistics (95%/95%, 2.Oxcy) +/- 0.0014 Fuel Eccentricity Included Basket Tolerances + 0.0039 Fuel Tolerances +/- 0.0091 Depletion Uncertainty +0.0067 Statistical Combination of Uncertainties +/- 0.0154 Reference k1f (MCNP4a) 0.9038 Total Uncertainty (above) 0.0154 Temperature Bias 0.0033 IFBA Bias PROPRIETARX REMOVED Calculational Bias (see Appendix A) 0.0012 Maximum keff 0.9266 Regulatory Limiting kff 0.95 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-32 Rev. 1

Table 4.7.12 Summary of the Criticality Safety Analyses for Configuration 2, Accident Condition Design Basis Burnup at 4.95 wt% 23U 0.0 GWD/MTU Soluble Boron 600 ppm Uncertainties MCNP Bias Uncertainty (95%/95%) + 0.0090 CASMO Bias Uncertainty (95%/95%)

REMfOVED1 Calculational Statistics (95%/95%, 2.Oxa) +/- 0.0014 Fuel Eccentricity Included Basket Tolerances +0.0039 Fuel Tolerances +/- 0.0091 Depletion Uncertainty N/A Statistical Combination of Uncertainties +/- 0.0139 Reference kf (MCNP4a) 0.9189 Total Uncertainty (above) 0.0139 Temperature Bias 0.0033 Calculational Bias (see Appendix A) 0.0012 Maximum keff 0.9373 Regulatory Limiting keff 0.95 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-33 Rev. 1

Table 4.7.13 Effect of External Reflection Condition Calculated k~ff Difference to Design Basis Calculation Full External Water 0.9118 Reference Reflection (Design Basis)

Reflective Boundary 0.9118 +/-0.0000 Condition added External Water replaced by 0.9102 -0.0016 Void HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-34 Rev. 1

Table 4.7.14 Additional Safety Margin Case Design Basis Reduced Conservatisms Cooling Time 0 years 5 years Moderator Temperature 620 F 605 F B203 content in Burnable 18.1 % 12.5 %

Poison Rods Burnup at which Burnable 33.1 GWd/mtU 19 GWd/mtU Poison Rods are removed Calculated k~ff (4.95 wt%, 38 0.9118 0.8961 GWd/mtU)

Difference to Design Basis Reference 0.0157 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4-35 Rev. 1

4.8 Acceptability of Storing Indian Point Unit 3 Fuel in the Indian Point Unit 2 Pool After the transfer of fuel in the STC is completed, fuel from Unit 3 is stored temporarily in the Unit 2 pool. This section evaluates the acceptability of this condition, and concludes that no further analyses are needed.

From a criticality perspective, the fuel from Indian Point Unit 2 (IP2) and Indian Point Unit 3 (IP3) are essentially the same. The most significant difference between the fuel types in the two plants is that IP2 fuel has more grid straps than the similar model for IP3, which is a detail too fine to affect the criticality model. Every type of fuel used at IP3 (LOPAR, OFA, Vantage5,.

Vantage+, 15x15 Upgrade) has also been used at IP2 and is resident in the IP2 Spent Fuel Pit (SFP). Otherwise, the pertinent physical characteristics of the fuel types, such as general physical geometry, uranium enrichment and loading are identical. Also, the use of burnable poisons, operating temperature and reactor power is comparably the same. Note that thermal power has increased gradually for both plants since initial operation.

~PROPRIETARY TEXT KEýM0VED.

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Appendix 4.A Benchmark Calculations (total number of pages: 17 including this page)

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4.A.1 INTRODUCTION AND

SUMMARY

Benchmark calculations have been made on selected U0 2 critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections.

MCNP4a [4.A. 1] is a continuous energy Monte Carlo code and KENO5a [4.A.2] uses group-dependent cross sections. For the KENO5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II [4.A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errorst (trends) that have been reported (e.g.,

[4.A.3 through 4.A.5]) for calculations with collapsed cross section sets.

In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the l0 B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.

Table 4.A.1 summarizes the benchmark experiments chosen to be modeled, the experimental uncertainty of these experiments and the results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.

One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KENO5a computes and prints the "energy of the average lethargy causing fission". In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KENO5a, the number-of fissions in each group may be collected and the energy of the average lethargy causing fission determined (post-processing).

Figures 4.A.1 and 4.A.2 show the calculated keff for the benchmark critical experiments as a function of the "energy of the average lethargy causing fission" for MCNP4a and KENO5a, respectively (U0 2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental error" in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.

Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.

t A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.

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Linear regression analysis of the data in Figures 4.A.1 and 4.A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KENO5a). The total bias (systematic error, or mean of the deviation from a klff of exactly 1.000) for the two methods of analysis are shown in the table below.

Calculational Bias of MCNP4a and KENO5a Total MCNP4a 0.0012 +/- 0.0090 KENO5a 0.0032 + 0.0113 The bias and standard error of the bias were calculated by the following equations from NUREG/CR-6698 [4.A. 19], with the standard error multiplied by the one-sided K-factor for 95%

probability at the 95% confidence level from NBS Handbook 91 [4.A.18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2).

1k n 1 (4.A.1)

Z 1 n

-2 f I1 k 1 n- I (.4.A .2) n j=1 o' i

--2 n (4.A.3) o--

y 2 i=1 (Ti Sp +/-+ (4.A.4)

Where:

ki is the calculated reactivity for the ith critical experiment, ai is the combination of the experimental and calculational uncertainty, determined by the square root of the sum of the squares, n is the number of critical experiments, kis the weighted mean keff value, HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS CONTAIN HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4.A-3 Rev. 1

s2 is the variance about the mean, a- is the average total uncertainty, SP is the pooled variance, Formulas 4.A.1 through 4.A.4 are based on the methodology of NUREG/CR-6698 [4.A.19].

Experimental uncertainty is combined with the calculational uncertainty by applying the square root of the sum of the squares of the individual uncertainties. (1-k) is the actual bias which is added to the MCNP4a and KENO5a results. Sp, is the uncertainty or standard error associated with the bias (based on the square root of the pooled variance). The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95% probability at the 95% confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively.

The bias values are used to evaluate the maximum klff values for the rack designs. KENO5a has a slightly larger systematic error than MCNP4a, but both result in greater precision than published data [4.A.3 through 4.A.5] would indicate for collapsed cross section sets in KENO5a (SCALE) calculations 4.A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46% to 5.74% and therefore span the enrichment range for rack designs. Figures 4.A.3 and 4.A.4 show the calculated keff values (Table 4.A.1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO5a). Thus, there are no corrections to the bias for the various enrichments.

As further confirmation of the absence of any trends with enrichment, a typical BWR high-density storage rack configuration was calculated with both MCNP4a and KENO5a for various enrichments. The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison between codes, shown in Table 4.A.2 and Figure 4.A.5, confirms that no significant difference in the calculated values of klff for the two independent codes as evidenced by the 450 slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.

4.A.3 Effect of 1°B Loading Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral/Metamic panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS CONTAIN HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4.A-4 Rev. I

measurements with absorber plates, but, with one exception (a flux-trap experiment), the reactivity worth of the absorbers in the PNL tests is very low and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.

Table 4.A.3 lists the subset of experiments using thin neutron absorbers (from Table 4.A. 1) and shows the reactivity worth (Ak) of the absorbernt No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 4.A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.

To further confirm the absence of a significant trend with 10B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO5a (as suggested in Reg. Guide 3.41).

Results are shown in Figure 4.A.6 and Table 4.A.4 for a typical BWR high-density storage rack geometry. These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 450 line, within an expected 95% probability limit).

4.A.4 Miscellaneous and Minor Parameters 4.A.4.1 Reflector Material and Spacings PNL has performed a number of critical experiments with thick steel and lead reflectors.t Analysis of these critical experiments are listed in Table 4.A.5 (subset of data in Table 4.A.1).

There appears to be a small tendency toward overprediction of kff at the lower spacing of the fuel rods from the reflector, although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.

4.A.4.2 Fuel Pellet Diameter and Lattice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3835 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.498 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel.

Thus, the critical experiments analyzed provide a reasonable representation of the fuel in the storage rack designs. Based on the data in Table 4.A. 1, there does not appear to be any tThe reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.

tParallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design. A lead reflector is also not directly pertinent, but might be used in future designs.

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observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments or the rack designs.

4.A.4.3 Soluble Boron Concentration Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO5a) calculations are shown in Table 4.A.6. Analyses of the very high boron concentration experiments (>1300 ppm) show a tendency to slightly over predict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.

4.A.4.4 Area of Applicability To ensure that the key physical parameters that define the system are appropriately covered by the selected benchmark calculations, the area of applicability must be defined. Table 4.A.8 provides a comparison of the key physical parameters in the benchmarks and typical storage rack configurations.

4.A.5 MOX Fuel The number of critical experiments with PuO 2 bearing fuel (MOX) is more limited than for U0 2 fuel. However, a number of MOX critical experiments have been analyzed and the results are also shown in Table 4.A.7. Results of these analyses are generally above a keff of 1.00, indicating that when Pu is present, both MCNP4a and KENO5a over predict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KENO5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO5a. It is also possible that the over prediction in kff for both codes may be due to a small inadequacy in the determination of the 24 1 Pu decay and 24 1Am growth. This possibility is supported by the consistency in calculated keff over a wide range of the spectral index (energy of the average lethargy causing fission).

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4.A.6 References

[4.A.1] J.F. Briesmeister, Ed., "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).

[4.A.2] SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation",

NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.

[4.A.3] M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460) Oak Ridge National Laboratory, September 1994.

[4.A.4] W.C. Jordan et al., "Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Systems, Inc., Oak Ridge National Laboratory, December 1986.

[4.A.5] O.W. Hermann et al., "Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.

[4.A.6] R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall, 1986.

[4.A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.

[4.A.8] G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW-1645-4, Babcock & Wilcox Company, November 1991.

[4.A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW- 1810, Babcock and Wilcox Company, April 1984.

[4.A. 10] J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75% Enriched Uranium-Oxide Rods," Trans. Am.

Nucl. Soc. 33: 362-364 (1979).

[4.A. 1 ] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 wt % and 4.31 wt % 2 35U Enriched HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS CONTAIN HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4.A-7 Rev. 1

U0 2 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.

[4.A. 12] S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 Wt% and 4.31 Wt% 235U Enriched UO 2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.

[4.A.13] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 Wt % 235U Enriched U0 2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.

[4.A. 14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.

[4.A.15] B.M. Durst et al., Critical Experiments with 4.31 wt % 235U Enriched UO 2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.

[4.A. 16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1986.

[4.A.17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.

[4.A. 18] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

[4.A. 19] Guide for Validation of Nuclear Criticality Safety Calcualtional Methodology, NUREG/CR-6698, U.S. Nuclear Regulatory Commission, January 2001.

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Table 4.A. 1 Summary of Criticality Benchmark Calculations 11OREAYIAIR%0L)

Notes: NC stands for not calculated.

t EALF is the energy of the average lethargy causing fission tt The experimental results appear to be statistical outliers (>3cy) suggesting the possibility of unusually large experimental error. Although they could be justifiably excluded, for conservatism, they were retained in determining the calculational basis.

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Table 4.A.2 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS PROPR`iEtARY'TEXT'REMOV()%ED t Based on the GE 8x8R fuel assembly.

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Table 4.A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS PROPRIFjTARY TE- RFMOVED.

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Table 4.A.4 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIESt FOR VARIOUS BORON LOADINGS PR~OLRIFT.UYTXT REMOVFD.

Based on 4.5% enrichment GE 8xSR fuel assembly.

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Table 4.A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORSt PROPR IFTA7 :FA*V FMVEDE>V

yt.I "S

t Arranged in order of increasing reflector fuel spacing.

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Table 4.A.6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS TARYJEXT MOVED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS CONTAIN HOLTEC PROPRIETARY INFORMATION REPORT 111-2094289 4.A-14 Rev. 1

Table 4.A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL NeROPRJE CstnsfrXTREMt alED.a Note: NC stands for not calculated HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS CONTAIN HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4.A-15 Rev. I

Table 4.A.8 COMPARISON OF KEY PARAMETERS Parameter Critical Experiments Analyzed Systems Materials 235u 235u Fissionable material 235 Uranium enrichment (wt% U), 2.46 - 5.74 < 5.0 Physical Form U0 2 U0 2 Fuel density (g/cc) 9.20- 10.40 10.0-10.7 Moderator Material H 20 H20 Temperature (K) 300 300 Reflector Material Material Lead, Steel Steel Neutron Absorbers Material Soluble boron, B-10 Fixed Soluble boron, B-10 Fixed neutron absorbers neutron absorbers Soluble boron amount (ppm) 0-1899 0-1200 Fixed neutron absorber Borated SS, B4C-Al B 4C-Al Geometry Shape Square Square Absorber worth (Ak) 0.01-0.19 -. 0.12 Fuel pellet diameter (inches) 0.311 - 0.444 0.303 - 0.3835 (PWR fuel) 0.3224 - 0.498 (BWR fuel)

Fuel rod pitch (inches) 0.476- 1.00 0.496 - 0.580 (PWR fuel) 0.488 - 0.740 (BWR fuel)

Neutron Energy Thermal Thermal HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS CONTAIN HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 4.A-16 Rev. I

Figures 4A. 1 through 4A.6 (FIGURES ARE WITHHELD IN ACCORDANCE WITH 10 CFR 2.390)

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APPENDIX 4.B:

VALIDATION OF CASMO-4 (APPENDIX WITHHELD IN ACCORDANCE WITH 10CFR2.390)

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CHAPTER 5: THERMAL-HYDRAULIC EVALUATION 5.0 Overview In this chapter the thermal-hydraulic adequacy of the Shielded Transfer Canister (STC) designed for onsite transfer of Indian Point Unit 3 (IP-3) fuel to the Indian Point Unit 2 (IP-2) is evaluated. The STC is a thick-walled vessel containing a fuel basket. The fuel basket design is similar to the Holtec Multi-Purpose Canisters (MPCs) deployed for storing fuel in the HI-STORM system [L.E] but substantially thicker in cross section to provide enhanced radiation protection. Up to twelve fuel assemblies can be accommodated in the STC fuel basket. For additional shielding the STC is emplaced in a HI-TRAC steel-lead-steel transfer cask having a thick bolted lid prior to on-site movement. To minimize fuel and cask temperatures the STC cavity, HI-TRAC annular space between STC and HI-TRAC, and the HI-TRAC waterjacket are filled with water. Cutaway views of the STC and HI-TRAC transfer cask are depicted in Chapter 1, Figures 1.3.1 and 1.3.2. The transfer process is described in Chapter 1. The thermal analyses consider passive rejection of decay heat from the Spent Nuclear Fuel (SNF) to the environment. The following scenarios are evaluated:

i. Evaluation of normal onsite transfer of IP-3 fuel.

ii. A (non-mechanistic) postulated accident event resulting in the rupture of the HI-TRAC water jacket.

iii. A (non-mechanistic) postulated 50-gallon transporter gas tank rupture and fire accident.

The STC thermal design is required to comply with the temperature limits of SFST-ISG-1 1 [E.K] to ensure fuel integrity, HI-STORM System temperature limits [L.E] to ensure cask integrity and vessel pressure limits. The thermal criteria are set forth in Chapter 3, Tables 3.1.1 and 3.2.1. The maximum permissible heat load is specified in Table 5.0.1.

.HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-1 Rev. I

Table 5.0.1 BOUNDING SHIELDED TRANSFER CASK THERMAL LOAD Condition Value Maximum Fuel Assembly Decay Heat 1105.2 W (four interior cells) 650 W (eight peripheral cells)

STC Decay HeatNote I 9.621 kW Ambient Temperature 100IF Insolation 10CFR71 solar flux (See Table 5.0.2)

Note 1: Prior to fuel transfer the plant operations must verify that the thermal payload in the STC will not result in exceeding IP-2 fuel pool decay heat limits.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-2 Rev. I

Table 5.0.2 10CFR71 INSOLATION DATA Surface Type 12-Hour Insolation (W/m2 )

Horizontally Transported Flat Surfaces Base None Other Surfaces 774.0 Non-Horizontal Flat Surfaces 193.5 Curved Surfaces 387.0 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-3 Rev. 1

5.1 'Thermal Design The on-site fuel transfer equipment consists of the STC situated inside a vertically oriented HI-TRAC transfer cask. The HI-TRAC transfer cask is equipped with a thick bolted lid to maximize shielding and physical protection. The SNF assemblies reside inside the STC, which is closed by a bolted lid. In this manner the fuel placed in the STC is protected by two rugged boundaries. The STC contains a stainless-steel honeycomb fuel basket with square-shaped compartments of appropriate dimensions to allow insertion of the fuel assemblies. The fuel basket panels are equipped with neutron absorbing panels sandwiched between a stainless steel sheathing plate and the fuel basket panel, along the entire length of the active fuel region. The STC is water filled to emulate the wet storage environment in the IP-2 and IP-3 fuel pools. In this manner fuel temperature excursions during loading, transfer and unload-to-pool are minimized.

The water in the STC cavity plays an important role in the STC thermal performance. The water fills the spaces between solid components and provides an improved conduction medium (compared to gases) for dissipating decay heat. Within the STC the water environment sustains a closed loop thermosiphon action, removing SNF heat by an upward flow of water through the storage cells.

Thermosiphon action is defined as buoyancy induced global circulation of water in the MPC. The thermosiphon action is pictorially illustrated in Figure 5.1.1. The STC is externally cooled by the water filled HI-TRAC annulus. The HI-TRAC annulus is cooled by the so-called "Rayleigh effect" defined as natural circulation in differentially heated cavities. The Rayleigh effect transports heat laterally across the HI-TRAC annulus. The heat reaching the HI-TRAC inner surface is transmitted laterally across the HI-TRAC steel-lead-steel body by conduction. The HI-TRAC body is water jacketed to provide neutron shielding. The water jacket dissipates heat from the HI-TRAC steel-lead-steel body by natural circulation of water in the jacket spaces. The HI-TRAC is externally cooled by radiation and natural convection heat dissipation to air.

An important thermal design criterion imposed on the STC is to limit the maximum fuel cladding temperature to within design basis limits (Chapter 3, Table 3.1.1). An equally important requirement is to minimize temperature gradients in the STC to minimize thermal stresses. In order to meet these design objectives, the STC basket is designed to possess certain distinctive characteristics, which are summarized in the following.

The STC design minimizes resistance to heat transfer within the basket and basket periphery regions.

This is ensured by an uninterrupted panel-to-panel connectivity realized in the all-welded honeycomb basket structure. The STC design incorporates top and bottom plenums with interconnected downcomer paths. The top plenum is formed by the gap between the bottom of the STC lid and the top of the honeycomb fuel basket, and by elongated semicircular holes in each basket cell wall. The bottom plenum is formed by semicircular holes at the base of all cell walls. The STC basket is designed to eliminate structural discontinuities (i.e., gaps) which introduce added thermal resistances to heat flow. Consequently, temperature gradients are minimized in the design, which results in lower thermal stresses within the basket. Low thermal stresses are also ensured by an STC design that permits unrestrained axial and radial growth of the basket.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT Hl-2094289 5-4 Rev. I

5.1.1 Over-Pressure Protection During fuel transfer operations the water inside the STC and HI-TRAC expands under heatup to normal operating temperatures. To protect the vessels from excessive hydraulic pressures air spaces are provided under the STC and HI-TRAC lids. The minimum heights of the air spaces are defined below:

STC Lid: 9 inches HI-TRAC Lid: 9.25 inches HOLTEC INTERNATIONAL COPYRIGHTED MATERIA' REPORT HI-2094289 5-5 Rev. 1

FIGURE 5.1.1: ILLUSTRATION OF FUEL BASKET COOLING BY THERMOSIPHON ACTION HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-6 Rev. 1

5.2 Thermal Properties of Materials Materials present in the STC are zircaloy, fuel (U0 2 ), carbon steel, stainless steel, METAMIC neutron absorber, lead, air and water. Materials present in the HI-TRAC transfer cask are carbon steel, lead, water. In Table 5.2.1, a summary of references used to obtain cask material properties for performing all thermal analyses is presented.

Tables 5.2.2 and 5.2.3 provide thermal conductivity of materials at several representative temperatures. Conductivity of METAMIC is provided in Table 5.2.8. Emissivity of key materials are provided in Table 5.2.4. The emissivity properties of painted external surfaces are generally excellent. Kern [R.E] reports an emissivity range of 0.8 to 0.98 for a wide variety of paints. In the STC thermal analysis, an emissivity of 0.85* is applied to painted surfaces.

In Table 5.2.5, the heat capacity and density of the STC and HI-TRAC transfer cask materials are presented. These properties are required in performing transient (i.e., hypothetical fire accident condition) analyses. The temperature-dependent viscosity of air is provided in Table 5.2.6.

Heat transfer from exposed cask surfaces is calculated by accounting for both natural convection and thermal radiation heat transfer. Natural convection is a monotonically rising function of the product of Grashof(Gr) and Prandtl (Pr) numbers. Following the approach developed by Jakob and Hawkins

[R.I], the product GrxPr is expressed as L3ATZ, where L is height of the exposed surface, AT is the outer surface temperature differential and Z is a parameter based on air properties, which are known functions of temperature, evaluated at the average film temperature. The temperature-dependent values of Z are provided in Table 5.2.7.

This is conservative with respect to prior cask industry practice, which has historically utilized higher emissivities [R.O].

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-7 Rev. 1

Table 5.2.1 THERMO-PHYSICAL PROPERTY REFERENCES Material Emissivity Conductivity Density Heat Capacity Air N/A Handbook [R.B] Ideal Gas Law Handbook [R.B]

Zircaloy [R.C], [R.P], NUREG Rust [R.D] Rust [R.D]

[R.Q], [R.G] [R.F]

U0 2 Note 1 NUREG Rust [R.D] Rust [R.D]

[R.F]

Stainless Steel Kern [R.E] ASME [R.H] Marks' [R.A] Marks' [R.A]

(machined forgings)*

Stainless Steel ORNL ASME [R.H] Marks' [R.A] Marks' [R.A]

Platest [R.K], [R.L]

Carbon Steel Kern [R.E] ASME [R.H] Marks' [R.A] Marks' [R.A]

Lead Note 1 Handbook [R.B] Handbook [R.B] Handbook [R.B]

Water Note 1 ASME [R.J] ASME [R.J] ASME [R.J]

METAMIC Note 1 Test Data Test Data Test Data

[R.M], [R.N] [R.M], [R.N] [R.M], [R.N]

Note 1: Emissivity not reported as radiation heat dissipation from these surfaces is conservatively neglected.

Used in the STC lid.

t Used in the basket panels, neutron absorber sheathing, STC shell and baseplate.

Rev. 1 HOLTEC INTERNATIONAL5-8 COPYRIGHTED MATERIAL HI-2094289 REPORT HI-2094289 5-8 Rev. I

Table 5.2.2 MATERIALS THERMAL CONDUCTIVITY DATA Material At 200OF At 450OF At 700°F At 1000 0F (Btu/ft-hr-°F) (Btu/ft-hr-*F) (Btu/ft-hr-°F) (Btu/ft-hr-°F)

Air* 0.0173 0.0225 0.0272 0.0336 Stainless Steel 8.4 9.8 11.0 12.4 Carbon Steel 24.4 23.9 22.4 20.0 Lead 19.4 17.9 16.9 N/A Water 0.392 0.368 N/A N/A

At lower temperatures, Air conductivity is between 0.0139 Btulft-hr-0 F at 32°F and 0.0176 Btu/ft-hr-°F at 212 0F.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-9 Rev. 1

Table 5.2.3 FUEL ASSEMBLY MATERIALS THERMAL CONDUCTIVITY DATA Zircaloy Cladding* Fuel (U0 2)

Temperature ('F) Conductivity Temperature (°F) Conductivity (Btu/ft-hr-°F) (Btu/ft-hr-°F) 392 8.28 100 3.48 572 8.76 448 3.48 752 9.60 570 3.24 932 10.44 793 2.28

Conductivities of other Zirconium based cladding materials such as Zirlo are well approximated by Zircaloy because the principal alloying element content is the same.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-10 Rev. I

Table 5.2.4 SURFACE EMISSIVITY DATA*

Material Emissivity Zircaloy 0.80 Painted surfaces 0.85 Stainless steel (machined 0.36 forgings)

Stainless Steel Plates 0.587**

Carbon Steel 0.66

See Table 5.2.1 for cited references.

- Lowerbound value from the cited references in Table 5.2.1.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-11 Rev. I

Table 5.2.5 DENSITY AND HEAT CAPACITY DATA Material Density (lbm/ft3) Heat Capacity (Btu/lbm-0 F)

Zircaloy 409 0.0728 Fuel (U0 2) 684 0.056 Carbon steel 489 0.1 Stainless steel 501 0.12 Lead 710 0.031

@ 170.3°F 60.8 Water -@ 260.3 °F 58.5 0.999

@ 350.3 °F 55.6

@ 440.3 OF 51.9 METAMIC 163.4** 0.22**

Air Ideal Gas 0.24 Lowerbound values reported for conservatism.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-12 Rev. I

Table 5.2.6 GASES VISCOSITY* VARIATION WITH TEMPERATURE Temperature ('F) Air Viscosity (Micropoise) 32.0 172.0 70.5 182.4 260.3 229.4 338.4 246.3 567.1 293.0 701.6 316.7 1078.2 377.6

Obtained from Rohsenow and Hartnett [R.B].

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-13 Rev. I

Table 5.2.7 VARIATION OF NATURAL CONVECTION PROPERTIES PARAMETER "Z" FOR AIR WITH TEMPERATURE Temperature (*F) Z (ft-3?Fl)*

40 2.1xl06 140 9.Ox105 240 4.6x1 0 340 2.6x10 5 440 1.5x105

Obtained from Jakob and Hawkins [R.I]

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-14 Rev. I

Table 5.2.8 METAMIC CONDUCTIVITY DATA (33 wt.% B4C)*

Temperature Conductivity

°C (OF) W/m-°C (Btu/ft-hr-°F) 25 (77) 98.4 (56.88) 100 (212) 98.0 (56.64) 250 (482) 101.6 (58.68)

For conservatism the B 4 C content is overstated.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-15 Rev. 1

5.3 Thermal Evaluation of Fuel Transfer Operation The STC basket is designed to accommodate up to twelve W-15xl5* IP-3 fuel assemblies. The fuel basket is a matrix of interconnected square compartments designed to hold the fuel assemblies in a vertical position during fuel transfer. The basket is a honeycomb structure of stainless steel plates with full-length edge-welded intersections to form an integral basket configuration. The cell walls are equipped with neutron absorber plates sandwiched between the box wall and a stainless steel sheathing plate over the full length of the active fuel region. The neutron absorber plates are made of aluminum and boron carbide-containing METAMIC Metal-Matrix-Composite material to provide criticality' control, while maximizing heat conduction capabilities.

Thermal analysis of the STC is performed under the bounding heat load scenarios defined in Table 5.0.1 wherein all fuel assemblies are assumed to be generating heat at the maximum permissible rate.

While the assumption of limiting heat generation in each storage cell imputes a certain symmetry to the cask thermal analysis, it grossly overstates the total heat duty because it is unlikely that any fuel basket would be loaded with fuel assemblies emitting heat at their limiting values. The principal attributes of the thermal model are described in the following:

i. Heat generation in the STC is axially non-uniform with peaking in the mid-section of the active fuel length.

ii. Inasmuch as the transfer of heat occurs from inside the basket region to the outside, the temperature field in the STC is spatially distributed with the maximum values reached in the central core region.

iii. Heat is dissipated in the fuel basket by internal convection of water (See Figure 5.1.1).

As the rate of heat transfer is a direct function of flow resistance, the thermal analysis is conservatively based on the assumption that all fuel storage locations are populated with the most resistive Westinghouse fuel, W-17x 17 fuel assemblies.

iv. Heat is dissipated from the external surfaces of the cask by radiation and natural convection to air.

5.3.1 Description of the 3-D Thermal Model

i. Introduction The STC interior is a 3-D array of square shaped cells inside an irregularly shaped basket outline confined inside the cylindrical space of the STC cavity. To ensure an adequate representation of these features, a 3-D geometric model of the STC is constructed using the FLUENT Computational Fluid Dynamics (CFD) code pre-processor [M.E]. Other than representing the composite cell walls (made up of stainless steel panels, neutron absorber panels and stainless steel sheathing) by a homogeneous panel with equivalent orthotropic (thru-thickness and parallel plates direction) thermal conductivities, the 3-D model requires no idealizations of the fuel basket structure. Further, since as

15x15 array Westinghouse fuel assemblies.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-16 Rev. 1

it is impractical to model every fuel rod in every stored fuel assembly explicitly, the cross section bounded by the inside of the storage cell, which surrounds the assemblage of fuel rods and the interstitial water (also called the "rodded region"), is replaced with an "equivalent" square homogeneous section characterized by an effective thermal conductivity. Homogenization of the storage cell cross-section is illustrated in Figure 5.3.1. For thermal-hydraulic simulation, each fuel assembly in its storage cell is represented by an equivalent porous medium.

ii. Details of the 3-D Model The 3-D model implemented has the following key attributes:

a. As mentioned above, the composite walls in the fuel basket consisting of the Alloy X* structural panels, the aluminum-based neutron absorber, and the Alloy X sheathing, are represented by an orthotropic homogeneous panel of equivalent thermal conductivity in the three principal directions. The in-plane and thru-thickness thermal conductivities of the composite wall are computed using a standard procedure for such shapes with certain conservatisms, as described below.

During fabrication, a uniform normal pressure is applied to each "Box Wall - Metamic -

Sheathing" sandwich in the assembly fixture during welding of the sheathing periphery on the box wall. This ensures adequate surface-to-surface contact between the neutron absorber and the adjacent Alloy X surfaces. The mean coefficient of linear expansion of the neutron absorber is higher than the thermal expansion coefficients of the basket and sheathing materials. Consequently, basket heat-up from the stored SNF will further ensure a tight fit of the neutron absorber plate in the sheathing-to-box pocket. Nevertheless the possible presence.

of small microscopic gaps due to less than perfect surface-to-surface contact requires consideration of an interfacial contact resistance between the neutron absorber and box-sheathing surfaces. In the thermal analysis a 2 mil neutron absorber to pocket gap has been used. This is conservative as the sandwich is engineered to ensure an essentially no-gap fitup and assembly of the neutron-absorber panels. Furthermore, no credit is taken for radiative heat exchange across the neutron absorber to sheathing or neutron absorber to box wall gaps.

The heat conduction properties of the composite "Box Wall - Metamic - Sheathing" sandwich panels in the two principal basket cross sectional directions (i.e., thru-thickness and parallel plates direction) are unequal. In the thru-thickness direction, heat is transported across layers of sheathing, water-gap, neutron absorber and box wall resistances that are essentially in series. Heat conduction in the parallel plates direction, in contrast, is through an array of essentially parallel resistances comprised of these several layers listed above. In this manner the composite walls of the fuel basket storage cells are replaced with a solid wall of equivalent through thickness and parallel plates direction conductivities.

Alloy X refers to the group of Stainless Steel grades 316, 316 LN, 304, and 304 LN permitted for basket construction. To ensure a bounding evaluation lowerbound stainless properties are defined in Section 5.2.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-17 Rev. I

b. The fuel storage spaces are replaced by an equivalent porous media having the flow impedance characteristics of the most resistive Westinghouse fuel, W-17x1 7 fuel assembly.

The flow resistance is obtained using a conservatively articulated 3-D CFD model [R.S]. As an additional measure of conservatism the guide tubes and instrument tubes are assumed to be plugged.

c. Every STC fuel storage cell is assumed to be occupied by fuel emitting heat at the maximum permissible rate (See Table 5.0.1). The in-plane thermal conductivity of the fuel assemblies are obtained from finite element models of an array of fuel rods enclosed by a square box. Heat transfer in the axial direction is computed by an area weighted mean of cladding and water conductivities. Axial conduction heat transfer in the fuel pellets is ignored.

d. The internals of the STC, including the basket cross section, cutouts at the bottom of the basket wall to allow water circulation, top plenum, and downcomer flow passages are modeled explicitly. For simplicity, the mouse holes are modeled as rectangular openings with understated flow area.

e. The HI-TRAC annulus, steel-lead-steel layers, top lid, pool lid and waterjacket are explicitly modeled.

The principal modeling conservatisms are listed below:

1) The storage cell spaces are loaded with the most resistive W-17x 17 fuel.

2) Fuel assembly guide tubes and instrument tubes are assumed to be blocked.

3) All storage cells are generating heat at their limiting value defined in Table 5.0.1.

4) Axial dissipation of heat by the fuel pellets is neglected.

5) The most severe environmental factors for long-term normal storage - ambient temperature of 100°F and 10CFR71 insolation levels (see Table 5.0.2)- were coincidentally imposed on the system.

6) No credit is taken for contact between fuel assemblies and the STC basket wall or between the STC basket and the basket supports.

7) Heat dissipation by fuel basket peripheral supports is neglected.

8) Cask surface emissivities are understated.

9) Cask bottom is assumed to be insulated.

10) Air motion in the STC and HI-TRAC cavities neglected.

5.3.2 Maximum Temperatures The 3-D model articulated in the previous subsection is used to determine temperature distributions under transfer of IP-3 fuel placed in the STC. The fuel transfer scenario assumes maximum permissible fuel heat load, hot ambient temperature (See Table 5.0.1), insolation heating and steady-maximum temperatures. The results of the analysis are tabulated in Tables 5.3.1 and 5.3.2. The HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-18 Rev. 1

following observations are derived by inspecting the temperature field obtained from the thermal models:

The fuel cladding temperatures are within the SFST-ISG-1 1 limits.

" The maximum temperature of the basket structural materials are within their design limits.

The maximum temperature of the neutron absorbers are within their design limits.

The maximum temperatures of the STC pressure boundary materials are within their design limits.

The maximum STC, HI-TRAC annulus and waterjacket pressures are within design limits (See Tables 3.1.1 and 3.2.1).

The above observations leads to the conclusion that the temperature field in the STC containing heat emitting IP-3 fuel provides a safe environment for stored fuel. The component temperatures and pressure boundary pressures are in compliance with thermal design criteria set forth in Chapter 3 and Section 5.3.

The STC and the HI-TRAC are in multiple configurations during the loading of the STC into the HI-TRAC during normal on-site transfer operations as described in Chapter 10. The temperatures reported in Table 5.3.1 bound the following loading configurations:

1. Loaded STC without the STC lid being sealed in the HI-TRAC.

2. Loaded STC with the STC lid sealed in the HI-TRAC without the HI-TRAC lid.

5.3.3 Evaluation of STC without the HI-TRAC An evaluation of the STC component temperatures was performed for a hypothetical situation wherein the STC is assumed to be suspended in air above the spent fuel pool. The evaluation assumes the following:

1. The STC is loaded with the fuel assemblies with the maximum permissible decay heat.

2. Heat transfer is conservatively assumed to occur only through the cylindrical surfaces of the STC.

3. The top surface of the STC lid and the bottom surface of the STC baseplate are assumed to be insulated.

4. The STC lid is in place.

5. An ambient temperature of 49 0 C (120 0 F) is assumed.

6. Heat transfer from the outside surfaces of the STC includes natural convection and surface-to-ambient thermal radiation.

7. Steady state maximum temperatures are reached.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-19 Rev. I

The evaluation shows that the average outer surface temperature of the STC without the HI-TRAC is 91 0 C (196 0 F). During on-site transfer operations, the average outer surface of STC placed within the HI-TRAC is 93 0 C (199 0 F). The bare (no HI-TRAC) STC components and cavity temperatures will, therefore, be lower than those reported in Table 5.3.1. This scenario bounds the following loading configurations:

1. Loaded STC with and without the STC lid.

2. Loaded STC without the STC lid being sealed in FSB between the spent fuel pool and the HI-TRAC.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-20 Rev. 1

Table 5.3.1 MAXIMUM TEMPERATURES UNDER FUEL TRANSFER Component Temperature*

°C (OF)

Fuel Cladding 105 (221)

Fuel Basket 105 (221)

STC Inner Shell 98 (208)

STC Closure Lid 92(198)

STC Lid Seal 92 ( 1 9 8 )Note I HI-TRAC Inner Shell 85 (185)

HI-TRAC Lid 75 (167)

HI-TRAC Lid Seal 75 (16 7 )Note I Water Jacket Shell 80 (176)

Water Jacket-Bulk 79 (174)

HI-TRAC Annulus Water Bulk 90 (194)

STC Water Bulk 103 (217)

Note 1: To bound the lid seal temperature the maximum lid temperature is tabulated herein.

The tabulated temperatures are within the thermal criteria set forth in Chapter 3, Table 3.1.1.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-21 Rev. I

Table 5.3.2 MAXIMUM PRESSURES UNDER FUEL TRANSFER Cavity Pressure* [psig]

STC 34.0 HI-TRAC 17.5

The tabulated pressures are within the structural criteria set forth in Chapter 3, Table 3.2.1.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-22 Rev. 1

HEATGENERATING BASKEMOCLLWALL FUaLRODS ARRAYI d

oo6obooooooooooo I 7 0000000000000000 0000000000000000 0000000000000000 0000000000000000 0000000000000000 0000000000000000 0000000000000000 1e000000000000000 WATER FILLING I' 0000000000000000 EMPTY SPACES 0000000000000000 0000000000000000 0000000000000000 0000000000000000 0000000000000000 0000000000000000 (a) TYPICAL FUEL CELL (b) SOUD REGION OF EFFECTIVE CONDUCTIVITY FIGURE 5.3. 1: HOMOGENIZATION OF THE STORAGE CELL CROSS-SECTION HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094299 5-23 Rev. I

5.4 Hypothetical Accident Evaluation To demonstrate the robustness of the STC fuel transfer operation severe accidents are postulated and evaluated herein. The accidents are defined as follows:

i. Postulated rupture of the HI-TRAC water jacket.

ii. Postulated 50-gallon transporter gas tank rupture and fire accident.

To ensure fuel and STC and HI-TRAC integrity the following criteria must be demonstrated:

The fuel cladding must remain below the SFST-ISG-1 1 temperature limit.

The STC vessel temperature and pressure must remain below accident limits.

HI-TRAC pressure boundary temperature and pressure must remain below accident limits.

The accidents are evaluated in Sections 5.4.1 and 5.4.2.

5.4.1 Jacket Water Loss Accident The integrity of fuel cladding STC pressure boundary integrity is evaluated under a postulated rupture and loss of water from the HI-TRAC water jacket. The HI-TRAC is equipped with an array of water compartments filled with water. For a bounding analysis, all water jacket compartments are assumed to be drained of water and replaced with air. The HI-TRAC is assumed to have the maximum thermal payload (Table 5.0.1) and assumed to have reached steady state maximum temperatures. Under this array of adverse conditions, the maximum temperatures and pressures are computed and reported in Tables 5.4.1 and 5.4.2. The results ofjacket water loss evaluation confirm that the cladding, STC and HI-TRAC component temperatures are below design limits and the co-incident STC pressure is bounded by the vessel design pressure.

5.4.2 Fire Accident Although the probability of a fire accident during fuel transfer operations is low, a conservative fire event has been assumed and analyzed. Entergy will implement administrative controls prior to each inter-unit transfer campaign to ensure there are no permanent or transient sources of fire in the vicinity of the transport path that create a condition outside the fire analysis and design basis of the HI-TRAC/STC assemblage. The fire event is defined as rupture of an on-site transport vehicle fuel tank filled to capacity and ignition of spilled fuel. The fuel tank capacity is limited to 50 gallons.

The fuel tank fire is conservatively assumed to surround the HI-TRAC in the manner described in Item 3 below. All exposed transfer cask surfaces are heated by radiation and convection heat transfer from the fire. Although not mandated by 10 CFR 50 Regulations, the NUREG- 1536 and 10 CFR 71 guidance is adopted to conservatively bound the consequences of the postulated fire event.

The fire parameters from the cited references are defined below:

1. The average flame emissivity is at least 0.9 and the cask absorbtivity at least 0.8.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-24 Rev. 1

2. The average flame temperature must be at least 1475'F (800'C). Open pool fires typically involve the entrainment of large amounts of air, resulting in lower flame temperatures.

Additionally, the bounding temperature is applied to all exposed cask surfaces, which is very conservative considering the size of the transfer cask. It is therefore conservative to use the 1475°F (800'C) temperature.

3. The fuel source must extend horizontally at least 1 m (40 in), but may not extend more than 3 m (10 ft), beyond the external surface of the cask. Use of the minimum ring width of 1 meter yields a deeper pool thereby conservatively maximizing the fire duration.

4. The convection coefficient must be that value which may be demonstrated to exist if the cask were exposed to the fire specified. Based on Sandia large pool fire thermal measurements

[R.R], a forced convection heat transfer coefficient of 4.5 Btu/(hrxft2 x°F) is applied to the exposed transfer cask surfaces during fire.

Based on the limiting 50 gallon fuel volume, the transfer cask diameter (7.4 ft) and the lowerbound 1 m fuel ring width, a fuel depth of 0.72 in is obtained. From this depth and lowerbound fuel consumption rate of 0.15 in/min, the fire duration is calculated to be 4.85 minutes. The fuel consumption rate of 0.15 in/min is the lowerbound value from Sandia large pool fire tests [R.R]. Use of a lowerbound fuel consumption rate conservatively maximizes the fire duration.

Based on the fire parameters defined by items 1 through 4 above the heat input to the HI-TRAC transfer cask is computed as follows:

qF = h&(TF - Ts) + MEat [(TF + C )4 - (Ts + C )4]

where:

qF = Cask Heat Flux (Btu/ft2 -hr) hf, = Forced Convection Heat Transfer Coefficient (4.5 Btu/ft2-hr-OF) a = Stefan-Boltzmann Constant TF = Fire Temperature (1475°F)

C= Conversion Constant (460 (°F to 'R))

Ts = Cask Surface Temperature (°F)

= Flame Emissivity (0.90)

= Cask Absorbtivity (0.8)

From the HI-TRAC fire analysis, a bounding rate of temperature rise (3.549°F per minute) is determined. The total temperature rise computed as the product of the rate of temperature rise and fire duration is 17'F. Applying this bounding temperature rise the temperature of the fuel and STC are obtained. The results are reported in Table 5.4.3. The coincident boundary pressures are computed and reported in Table 5.4.4. The following observations are derived from the fire accident results:

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-25 Rev. 1

" The fuel cladding temperature is within the SFST-ISG-11 limits.

" The maximum temperature of the basket structural materials are within design limits.

The maximum temperature of the METAMIC neutron absorber is within design limits.

" The maximum temperatures of the STC pressure boundary materials are within design limits.

The maximum STC and HI-TRAC pressures are within design limits..

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-26 Rev. 1

Table 5.4.1 JACKET WATER LOSS ACCIDENT TEMPERATURES Component Temperature*

0 C (OF)

Fuel Cladding 116 (241)

Fuel Basket 116(241)

STC Inner Shell 110 (230)

STC Closure Lid 104 (219)

STC Lid Seal 104 ( 2 1 9 )Note I HI-TRAC Inner Shell 100 (212),

HI-TRAC Lid 81 (178)

HI-TRAC Lid Seal 81 (17 8 )Note 1 Water Jacket Shell 88 (190)

HI-TRAC Annulus Water Bulk 103 (217)

STC Water Bulk 115 (239)

Note 1: To bound the lid seal temperature the maximum lid temperature is tabulated herein.

The tabulated temperatures are within the thermal criteria set forth in Chapter 3, Table 3.1.1.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT 1-1-2094289 5-27 Rev. 1

Table 5.4.2 JACKET WATER LOSS CAVITY PRESSURES Cavity Pressure* [psig]

STC 53.5.

HI-TRAC 25.5

The tabulated pressures are within the structural criteria set forth in Chapter 3, Table 3.2.1.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-28 Rev. 1

Table 5.4.3 FIRE ACCIDENT TEMPERATURES Component Temperature*

0C [(OF)]

Fuel Cladding 115 (239)

Fuel Basket 115 (239)

STC Inner Shell 108 (226)

STC Closure Lid 102 (216)

STC Lid Seal 102 (21 6 )Nole I HI-TRAC Inner Shell 95 (203)

HI-TRAC Lid 85 (185)

HI-TRAC Lid Seal 85 ( 1 8 5 )Note I Water Jacket Shell 90 (193)

Water Jacket Bulk 89 (191)

HI-TRAC Annulus Water Bulk 100 (212)

STC Water Bulk 113 (235)

Note 1: To bound the lid seal temperature the maximum lid temperature is tabulated herein.

Table 5.4.4 FIRE ACCIDENT CAVITY PRESSURES Cavity Pressuret [psig]

STC 5 3 .5Note 1 HI-TRAC 25.5 Note I Note 1: For conservatism the jacket water loss cavity pressures are adopted herein. These pressures bound the fire accident pressures because the jacket water loss temperatures exceed the fire accident temperatures (See temperature Tables 5.4.1 and 5.4.3).

The tabulated temperatures are within the thermal criteria set forth in Chapter 3, Table 3.1.1.

t The tabulated pressures are within the structural criteria set forth in Chapter 3, Table 3.2.1.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 5-29 Rev. 1

CHAPTER 6: STRUCTURAL EVALUATION OF NORMAL AND ACCIDENT CONDITION LOADINGS 6.0 Overview In this chapter, the structural components of the Shielded Transfer Canister (STC) are identified and described. The objective of the structural analyses is to ensure that the integrity of the STC and the HI-TRAC is maintained under normal, off-normal and extreme environmental conditions as well as all credible accident events. The results of the structural analyses, summarized in this chapter, support the conclusion that the STC and the HI-TRAC meet the structural design criteria set forth in Chapter 3. To facilitate regulatory review, the assumptions and conservatisms inherent in the analyses are identified along with a concise description of the analytical methods, models, and acceptance criteria.

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6.1 Structural Design 6.1.1 Discussion The STC consists of a fuel basket inside a thick-walled cylindrical vessel (Figure 1.3.1).

The HI-TRAC transfer cask is an existing piece of equipment at IP-2, which is licensed under NRC docket 72-1014 as part of the HI-STORM 100 Dry Cask Storage System. The HI-TRAC with its specially designed closure lid is shown in Figure 1.3.2. A complete description of the design details of the STC and the HI-TRAC is provided in Section 1.3.

In this section, the discussion is confined to characterizing the design features of the STC and the HI-TRAC transfer cask relevant to their structural analysis.

6.1.1.1 Shielded Transfer Canister (STC)

IkOPRIETA- TEXT REMOVED.

OEY As stated in Chapter 3, the STC is designed to meet ASME Code,Section III, Subsection ND stress limits.

6.1.1.2 HI-TRAC 1OOD Transfer Cask PROPRIETARY TEXT REMOVED 1 N The structural steel components of the HI-TRAC are subject to the stress limits of the ASME Code,Section III, Subsection ND, Class 3 for normal and off-normal loading conditions. The stress limits, for the HI-TRAC closure lid lifting, are conservatively set to follow guidelines from NUREG-0612.

6.1.2 Design Criteria and Applicable Loads Principal design criteria for the design basis, normal condition, and accident/environmental loads are discussed in Sections 3.2 and 3.3. In this section, the loads, load combinations, and the required structural performance of the STC and the HI-TRAC under the various loading events are presented.

Stresses arise in the components of the STC and the HI-TRAC due to various loads that originate under design, normal, or accident conditions. These individual loads are combined to form load combinations. Stresses, strains, displacements, and stress intensities, as applicable, resulting from the load combinations are compared to their respective allowable limits. The following subsections present loads, load combinations, and the allowable limits germane to them for use in the structural analyses of the STC and the HI-TRAC transfer cask.

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6.1.2.1 Loads and Load Combinations The individual loads applicable to the STC and the HI-TRAC cask are defined in Section 3.2 of this report. Load combinations are developed by assembling the individual loads that may act concurrently, and possibly, synergistically. The load combinations, which are summarized in Table 3.2.4, are applied to the mathematical models of the STC and the HI-TRAC. Results of the analyses carried out under bounding load combinations are compared with their respective allowable limits in Section 6.2. The analysis results from the bounding load combinations are also evaluated to ensure satisfaction of the functional performance criteria discussed in the foregoing.

6.1.2.2 Materials and Allowables All the load bearing members of the STC (viz. the shells, the closure lid, the baseplate and the upper flange) are built from SA-516 Gr. 70 or equivalent material. The STC trunnions and the closure lid bolts are made from SA 564 630 HI 100 material. The STC basket is made from Alloy X material. The material properties of Alloy X are the least favorable values from the set of candidate stainless steel types: 316, 316 LN, 304, and 304 LN. The standard HI-TRAC 100D top lid is replaced with a solid circular lid without any penetrations which is made of SA 516 Gr. 70 material. The HI-TRAC pool lid bolts, which are re-analyzed in this application due to the HI-TRAC internal pressure, are made of SA-193 B7. The detailed discussion about these structural materials can be found in Section 3.3 of the HI-STORM FSAR [K.A].

Allowable stresses and stress intensities are calculated using the data provided in the ASME Code [G.B]. Tables 6.1.1 through 6.1.5 contain numerical values of the stresses/stress intensities for all STC and HI-TRAC load bearing materials as a function of temperature.

In all tables the terms S, Sy, and S,, respectively, denote the design stress, minimum yield strength, and the ultimate strength. Property values at intermediate temperatures that are not reported in the ASME Code are obtained by linear interpolation. Property values are not extrapolated beyond the limits of the Code in any structural calculation.

Additional terms relevant to the analyses are extracted from the ASME Code (Section ND-3321) as follows:

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Symbol Definition Description Urn General membrane This stress is equal to the average stress across the solid stress section under consideration. It excludes discontinuties and concentrations, and is produced only by pressure and other mechanical loads GL Local membrane stress This stress is the same as cym, except that it includes the effect of discontinuties Ob Bending stress This stress is equal to the linear varying portion of the stress across the solid section under consideration. It excludes discontinues and concentrations, and is produced only by pressure and other mechanical loads.

S Allowable Stress Allowables stress value given in Table IA and lB.

It is recognized that the planar temperature distribution in the fuel basket and the STC under the maximum heat load condition is the highest at the canister center and drops monotonically, reaching its lowest value at the outside surface. Strictly speaking, the allowable stresses/stress intensities at any location in the STC should be based on the coincident metal temperature under the specific operating condition. However, in the interest of conservatism, design temperatures are established in Table 3.1.1 for each component, which are upper bounds on the metal temperature for each situational condition.

Finally, the lifting load path in the STC and the HI-TRAC transfer cask are subject to specific limits set forth by NUREG-0612 [C.A]. The following table summarizes the lift path and applicable guidance.

Component Applicable Allowable Stress Limit Guidance STC and HI-TRAC Trunnions In absence of the redundant load path the STC Lifting Points on the Lid NUREG-0612, induced stresses must be Section 5.1.6, (3) less the 1 / 1 0 th the ultimate HI-TRAC Lid Lifting Points strength of the applicable material.

However, for conservatism the primary stresses in STC and the HI-TRAC components in the lift load path are set to meet the smaller of 1/10 of the material ultimate strength and 1/6 of the material yield strength under a normal handling condition.

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TABLE 6.1.1 SA516 GRADE 70 MATERIAL PROPERTIES Temp. SA516, Grade 70 (Deg. F)

S, S, S E

-20 38.0 30.0 100 38.0 29.3 150 35.7 29.0 200 34.8 70.0 20.0 28.8 250 34.2 28.6 300 33.6 28.3 350 33.05 28.0 400 32.5 27.7 450 31.75 27.4 Definitions:

Sy = Yield Stress (ksi)

Su = Ultimate Stress (ksi)

S = Maximum Allowable Stress (ksi)

E = Young's Modulus (psi x 106)

Notes:

1. Source for Sy values is Table Y-1 of [G.B].

2. Source for Su values is Table U of [G.B].

3. Source for S values is Table IA of [G.B]

4. Source for E values is "Carbon steels with C less than or equal to 0.30%" in Table TM-1 of [G.B].

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TABLE 6.1.2 ALLOWABLE STRESS Code: ASME ND Material: SA-516, Grade 70 Classification and Value (ksi)

Service Temp (Deg. F) Maximum Membrane Membrane Condition Allowable Stress plus Bending Stress 6m Stress S (am or UL) + Gb Design and Level A -20 to 500 20 20 30 Level D 40 48 Notes:

1. S = Maximum allowable stress values from Table 1A of ASME Code,Section II, Part D.

2. Stress Limits per Table ND-3321-1.

TABLE 6.1.3 ALLOWABLE STRESS Code: ASME ND Bolt Material: SA-564, 630 Classification and Value (ksi)

Service Temp (Deg. F) Maximum Primary Stress Condition Allowable ar Stress S

Design and Level A 28 28 28 Level A -20 to 500 Level D 56 Notes:

1. S = Maximum allowable stress values from Table 3 of ASME Code,Section II, Part D.

2. Stress Limits per Table ND-3321-1.

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TABLE 6.1.4 ALLOWABLE STRESS Code: ASME ND Bolt Material: SA-193, B7 Classification and Value (ksi)

Service Temp (Deg. F) Maximum Primary Stress Condition Allowable am Stress S

Design and Level A -20 to 500 25 25 Level D 50 Notes:

1. S = Maximum allowable stress values from Table 3 of ASME Code,Section II, Part D.

2. Stress Limits per Table ND-3321-1.

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TABLE 6.1.5 ALLOY X MATERIAL PROPERTIES Alloy X r Temp.

(Deg. F) SY S, E

-40 30.0 75.0 28.82 100 30.0 75.0 28.14 150 27.5 73.0 27.87 200 25.0 71.0 27.6 250 23.75 68.5 27.3 300 22.5 66.0 27.0 350 21.6 65.2 26.75 400 20.7 64.4 26.5 Definitions:

Sy = Yield Stress (ksi) cc = Mean'Coefficient of thermal expansion (in./in. per degree F x 106)

S, = Ultimate Stress (ksi)

E = Young's Modulus (psi x 106)

Notes:

I. Source for Sy values is Table Y-1 of [G.B].

2. Source for S, values is Table U of [G.B].

3. Source for E values is material group G in Table TM-1 of [G.B].

Y Alloy X represents the least favorable strength value of all the alloys corresponding to SA-240 plate material (type 304, 304LN, 316, and 316LN).

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6.2 Structural Analysis Calculations of the stresses in the different components of the STC and the HI-TRAC from the effects of mechanical load case assembled in Table 3.2.4 are presented in the following.

The purpose of the analyses summarized herein is to provide the necessary assurance that there will be no unacceptable risk of loss to configuration assumed by criticality analysis, unacceptable release of radioactive material, unacceptable radiation levels, or impairment of ready retrievability of fuel from the STC and the STC from the HI-TRAC transfer cask.

Each load case in Table 3.2.4 is considered sequentially and all affected components are analyzed to determine the factors of safety.

6.2.1 Load Case 1: Design Pressure 6.2.1.1 STC Since the STC is a pressure vessel, calculations are performed to demonstrate that the stresses that develop in the STC shell, baseplate, and the closure lid under design internal pressure meet the ASME Subsection ND stress limits. The stresses in the STC shell ar-e calculated using the classical formula for thin-walled pressure vessels, which are:

a1 Pr

=--h'r=- Pr 2t t where P internal pressure (per Table 3.2.1);

r = mean radius of STC inner shell = 21.5 in; t = thickness of STC inner shell = 1 in.

The circumferential stress (Cyh), the axial stress (a7), and the radial stress (Cyr) are computed for both normal and accident internal pressures. The results are given in the following table:

Pressure Gh (psi) (71 (psi) Cyr (psi)

Normal, P = 50 psi 1,075 537.5 -50 Accident, P = 65 psi 1,397.5 698.75 -65 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-9 Rev. 1

It is noted that the above calculations conservatively assume that the STC inner shell is acting alone to resist the internal pressure, without any credit for the strengthening effects of the lead backing or the STC outer shell. Tables 6.1.2 and 6.1.3 provide the allowable membrane strength for Level A and Level D conditions. A safety factor greater than 1.0 exists for the case of normal and accident pressures.

The STC closure lid is modeled as a simply supported plate and is subject to the design internal pressure. The radius of the plate is set to be equal to the bolt circle diameter, and the thickness of the plate is equal to the minimum closure lid thickness.

The bolts are evaluated for the cumulative load from STC internal pressure and the load to provide a leak tight seal under normal and hypothetical accidental loadings. The bolts shall be prestressed to the corresponding maximum resultant load.

The results are reported in the table below:

Maximum Allowable Component Pressure Stress (psi) Stress (psi) Safety Factor Lid P = 65 psi 1,752 30,000 17.1 (Table 3.2.1)

Bolts 24,380 28,000 1.15 Note: Conservatively Level A stress limits are used under the accidental pressure load.

The stresses in the STC baseplate due to design internal pressure are bounded by the results in Subsection 6.2.3.2, which considers the combined effects of internal pressure plus normal handling.

6.2.1.2 HI-TRAC The HI-TRAC is also analyzed for the design pressures in Table 3.2.1. The circumferential stress (Ch), the axial stress (al), and the radial stress (Or) are computed for both normal and accident internal pressures. The stresses in the HI-TRAC shell are calculated using the classical formula for thin-walled pressure vessels, which are:

cr1 =--

Pr Pr_ a -

2t h t where P = internal pressure (per Table 3.2.1);

r = mean radius of HI-TRAC inner shell = 34.38 in; t = thickness of HI-TRAC inner shell = 0.75 in.

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The results are given in the following table:

Pressure Cyh (psi) CYl(psi) aYr (psi)

Normal, P = 30 psi 1,375.2 687.6 -30 Accident, P = 40 psi 1,833.6 916.8 -40 It is noted that the above calculations conservatively assume that the HI-TRAC inner shell is acting alone to resist the internal pressure, without any credit for the strengthening effects of the lead backing or the HI-TRAC outer shell. Tables 6.1.2 and 6.1.3 provide the allowable membrane strength for Level A conditions. A safety factor greater than 1.0 exists for the case of normal and accident pressures.

The HI-TRAC top lid is modeled as a simply supported plate and is subject to the design internal pressure. The radius of the plate is set to be equal to the bolt circle diameter, and the thickness of the plate is equal to the minimum top lid thickness.

The bolts are evaluated for the cumulative load from HI-TRAC internal pressure and the load to provide a leak tight seal under normal and hypothetical accidental loadings. The bolts shall be prestressed to the corresponding maximum resultant load.

The results are reported in the table below:

Maximum Allowable Component Pressure Stress (psi) Stress (psi) Safety Factor Top Lid P = 40 psi 16,660 30,000 1.8 (Table 3.2.1)

Top Lid Bolts 12,630 25,000 1.98 Note: Conservatively Level A stress limit are used under the accidental pressure load.

The stresses in the HI-TRAC pool lid due to design internal pressure are bounded by the results in Subsection 6.2.3.4, which considers the combined effects of internal pressure plus normal handling.

6.2.2 Load Case 2: Normal Operating Pressure Plus Temperature There are no significant thermal stresses in STC enclosure vessel since the presence of water both inside and outside of the STC minimizes the thermal gradients across the pressure boundary.

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The stress calculations in Subsection 6.2.1 for Load Case 1 bound the results for this load case since (i) the design internal pressure considered in Subsection 6.2.1 bounds the normal operating pressure, and (ii) the allowable stresses used in Subsection 6.2.1 are based on the temperature limits in Table 3.1.1 for normal operation.

6.2.3 Load Case 3: Normal Handling In this subsection, analyses for all lifting operations applicable to the transfer of fuel from IP-3 to IP-2 using the STC and the HI-TRAC are presented to demonstrate compliance with applicable codes and standards.

The following components participate in lifting operations: lifting trunnions located at the top of the HI-TRAC transfer cask, lifting trunnions and closure lid located at the top of the STC, STC baseplate, HI-TRAC pool lid, and lid lifting connections for the HI-TRAC closure lid and STC closure lid.

The HI-TRAC 100)D lifting trunnions and the surrounding structure are analyzed in Section 3.4.3 of the HI-STORM FSAR [K.A] for a bounding lifted weight of 200,000 lb (as compared to a total weight of 190,000 lb. for the HI-TRAC 100D including a fully loaded STC).

The evaluation of the adequacy of the participating components entails careful consideration of the applied loading and associated stress limits. The load combination D

+ H, where H is the "handling load", is the generic case for all lifting adequacy assessments. The .term D denotes the dead load. Quite obviously, D must be taken as the bounding value of the dead load of the component being lifted. In all lifting analyses considered in this document, the handling load H is assumed to be 0.151D. In other words, the inertia amplifier during the lifting operation is assumed to be equal to 0.15g. This value is consistent with the guidelines of the Crane Manufacturer's Association of America (CMAA) [N.A], Specification No. 70, 1988, Section 3.3, which stipulates a dynamic factor equal to 0.15 for slowly executed lifts. Thus, the "apparent dead load" of the component for stress analysis purposes is D* = 1. 15D. Unless otherwise stated, all lifting analyses in this report use the "apparent dead load", D*, as the lifted load.

In general, the stress analysis to establish safety pursuant to NUREG-0612, Regulatory Guide 3.61, and the ASME Code, requires evaluation of three discrete zones which may be referred to as (i) the trunnions and other lift points, (ii) the trunnion/component interface, hereinafter referred to as Region A, and (iii) the rest of the component, specifically the stressed metal zone adjacent to Region A, herein referred to as Region B.

Stress limits germane to each of the above three areas are discussed below:

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i. Trunnions and other lift points: NUREG-0612 recommends that under the "apparent dead load", D*, the maximum primary stress in the trunnion and the other lifting point be less than 10% of the trunnion material ultimate strength. However, for conservatism the recommendations from ANSI N14.6 [B.S] are also implemented by considering the additional stress limit of 1 /6 th the material yield strength.

ii. Region A: Trunnion/Component Interface: Stresses in Region A must meet ASME Code Level A limits under applied load D*. Additionally, for conservatism the recommendations from Regulatory Guide 3.61 are implemented to show that the primary stress in the cross-section does not exceed yield strength of the applicable material under load 3D*.

iii. Region B: Typically, the stresses in the component in the vicinity of the trunnion/component interface are higher than elsewhere. However, exceptional situations exist. For example, when lifting a loaded STC, the STC baseplate, which supports the entire weight of the fuel and the fuel basket, is a candidate location for high stress even though it is far removed from the lifting location. Even though the STC baseplate would normally belong to the Region B category, for conservatism it is considered as Region A in this report. The pool lid of the HI-TRAC transfer cask also fall into this dual category. In general, however, all locations of high stress in the component under D* must also be checked for compliance with ASME Code Level A stress limits.

Unless explicitly stated otherwise, all analyses of lifting operations presented in this report follow the load definition and allowable stress provisions of the foregoing.

Consistent with the practice adopted throughout this chapter, results are presented in dimensionless form, as safety factors, defined as Safety Factor,8 =-Allowable Stress in the Region Considered Computed Maximum Stress in the Region The safety factor, defined in the manner of the above, is the added margin over what is recommended by the applicable code (NUREG-0612 or ASME or Regulatory Guide 3.61).

In the following subsections, each of the lifting analyses performed to demonstrate compliance with regulations are briefly described. Summary results are presented for each of the analyses.

It is recognized that stresses in Region A are subject to two distinct criteria, namely Level A stress limits under D* and yield strength at 3D*. The applicable criteria is identified in the summary tables, under the column heading "Item", using the "3D*" identifier.

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All of the lifting analyses reported on in this Subsection are designated as Load Case 3 in Table 3.2.4.

6.2.3.1 STC Lifting Trunnions For the analysis of the trunnion, an accepted conservative technique for computing the bending stress is to assume that the lifting force is applied at the tip of the trunnion "cantilever" and that the stress state is fully developed at the base of the cantilever. This conservative technique, recommended in NUREG-1536, is applied to the STC lifting trunnions analyzed in this report.

The two lifting trunnions for the STC are spaced at 180 degrees. The trunnions are designed for a two-point lift in accordance with the aforementioned NUREG-0612 criteria. The lifting analysis demonstrates that the stresses in the trunnions, computed using the conservative methodology described above, comply with NUREG-0612 provisions.

Specifically, the following results are obtained:

STC Lifting Trunnionst Value (psi) Safety Factor Bending stress 15,011 6.78 Shear stress 6,788 9 t The lifted load is 200,000 lbf (bounds the actual lifted weight from the pool)

Note that the safety factor presented in the previous table represents the additional margin beyond the mandated limit of 6 on yield strength and 10 on tensile strength.

6.2.3.2 STC Lifting PQPwrliElTAYmJE RE- . ::VED HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-14 Rev. 1

Summary of Results from STC Lifting Induced Stress Allowable Stress Minimum Item (or Max. Load) (or Capacity) Safety Factor STC Closure Lid (psi) 6,928 69830,000 4.33 Lid Lifting Points (lbf)

____________ 92,000 169,000 1.84 6.2.3.3 STC Baseplate 1jPRO0PR1E'T'ARWITEXT REMOý'vl ED __).>

Summary of Results for STC Baseplate under Normal Handling (Load Case 3)

Item Value Allowable Safety Factor Bending Stress in Baseplate (psi) 1,545 30,000 19.4 Bending Stress in Baseplate (3D*) (psi) 2,665 33,600 12.6 6.2.3.4 HI-TRAC Pool (Bottom) Lid Section 1.5 of the FSAR [K.A] lists various drawings for the design and construction of the HI-TRAC. Specifically, Holtec dwg. 2145 and 4128 pertain to the HI-TRAC 100D.

During lifting of the HI-TRAC, the HI-TRAC pool lid supports the weight of a loaded STC plus water. Calculations are performed to show structural integrity of the HI-TRAC pool lid under this condition. In accordance with the general guidelines set down at the beginning of Subsection 6.2.3, the pool lid is considered as Region A for evaluating safety factors. The analysis shows that the stress in the pool lid is less than the Level A allowable stress under pressure equivalent to the heaviest STC, contained water, and lid self weight. Stresses in the lids and bolts are also shown to be below yield under three times the applied lifted load (using Regulatory Guide 3.61 criteria). The threaded holes in the HI-TRAC pool lid are also examined for acceptable engagement length under the condition of lifting the STC. It is demonstrated that the pool lid peripheral bolts have adequate engagement length into the pool lid to permit the transfer of the required load.

The safety factor is defined based on the strength limits imposed by Regulatory Guide 3.61.

The following table summarizes the results of the analyses for the HI-TRAC pool lid.

Results given in the following table compare calculated stress (or load) and allowable HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-15 Rev. 1

stress (or load). In all cases, the safety factor is defined as the allowable value divided by the calculated value.

Summary of Results for HI-TRAC Pool Lid under Normal Handling (Load Case 3)

Item Value Allowable Safety Factor Bending Stress in Pool Lid Top Plate (psi) 27,160 30,000 1.1 Bending Stress in Pool Lid Bottom Plate (psi) 6,789 30,000 4.4 Pool Lid Bolt (psi) 18,380 25,000 1.4 Bending Stress in Pool Lid Top Plate (3D*) (psi) 30,190 32,500 1.1 Bending Stress in Pool Lid Bottom Plate (3D*) 7,547 32,500 4.3 (psi)

Pool Lid Bolt Force (kips) (3D*) 20,430 91,500: 4.5 6.2.3.5 Lid Lifting Analyses The STC lid lifting analysis is performed to ensure that the threaded connections provided in the lid are adequately sized. The lifting analysis of the STC closure lid is based on a vertical orientation of loading from an attached lifting device.

In addition to the STC closure lid lifting analysis, the strength qualification of the lifting holes for the HI-TRAC top lid has been performed. The qualification is based on the NUREG-0612 for a non-redundant lifting system. Loads to lifting devices are permitted to be at a maximum angle of 45 degrees from vertical. A summary of results, pertaining to the various lid lifting operations, is given in the table below:

Summary of Lid Lifting Analyses Item Dead Load (lb) Minimum Safety Factor STC Closure Lid 4500 24.95 HI-TRAC Top Lid 5000 3.4 The analysis also demonstrates that thread engagement is sufficient for the threaded holes used solely for lid lifting.

6.2.4 Load Case 4: Fuel Assembly Drop Accident PROPRIETA~RY TEXT RZEMOVED.

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6.2.5 Load Case 5: HI-TRAC Vertical Drop Accident HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SIHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-17 Rev. 1

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6.2.6 Load Cases 6 and 7: Seismic Stability of Loaded VCT and Loaded HI-TRAC PROPRIETARY [TEXT ,REMOVE[).

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6.2.7 Load Case 8: Seismic Stability of STC in the Fuel Pool

~PROPHUITTARY, TEX T RIMOVED.7 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-20 Rev. 1

6.3 Figures Figure 6.2.1 Finite Element Model Set-up for the STC Lifting HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-21 Rev. 1

U Figure 6.2.2 Resulting Stress from STC Lifting I 1-HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-22 Rev. 1

,z Figure 6.2.3 Finite Element Model for the HI-TRAC 1OOD Drop Analysis HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-23 Rev. 1

Mat Id Al1 0-S > -40o

0 N S-60 Time (sec)

Figure 6.2.4 HI-TRAC 100)D Transfer Cask Impact Velocity Time History HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-24 Rev. 1

Mat Id Al1

° 30 o

"U

"'00

-10 0

rin=-1.0234 m =*A1 1.02 Time (sec)

Figure 6.2.5 Deceleration Time History of the Dropped HI-TRAC IOOD Transfer Cask HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 6-25 Rev. 1

CHAPTER 7: SHIELDING DESIGN AND ALARA CONSIDERATIONS

7.0 INTRODUCTION

This chapter presents the shielding design evaluation of the shielded transfer canister (STC) used to transfer fuel assemblies from the Indian Point Unit 3 spent fuel pool (SFP) to the Indian Point Unit 2 SFP. The STC will be placed inside a HI-TRAC for the transfer between Unit 3 and Unit 2.

Therefore, dose rates are provided for the STC by itself, as well as dose rates for the STC inside a HI-TRAC. In addition, ALARA considerations for occupational exposures received while transferring the STC into the HI-TRAC, from Unit 3 to Unit 2, and unloading the STC in Unit 2 are evaluated. Further, site boundary dose rates from transferring the HI-TRAC containing the STC between Unit 3 and Unit 2 for normal (10 CFR 72.104) and accident conditions (10 CFR 72.106) are provided in this chapter. The following information is included in this chapter:

A description of the shielding features of the STC and acceptance criteria.

A description of the source terms.

A general description of the shielding analysis methodology.

" A description of the analysis assumptions and results for the STC and HI-TRAC evaluations.

Analyses of site boundary dose rates for normal (10 CFR 72.104) and accident conditions (10 CFR 72.106).

Evaluation of occupational exposures per the ALARA principles in accordance with 10CFR20.1 101 (b).

" Evaluation of an effluent dose release during normal conditions.

The principal sources of radiation in the STC are:

" Gamma radiation originating from the following sources

1. Decay of radioactive fission products

2. Hardware activation products generated during core operations 3, Secondary photons from neutron capture in fissile and non-fissile nuclides

Neutron radiation originating from the following sources

1. Spontaneous fission

2. ax,n reactions in fuel materials

3. Secondary neutrons produced by fission from subcritical multiplication Shielding from gamma radiation is provided by the steel and lead shielding structures of the STC and HI-TRAC. In the STC design, Metamic is used in the basket structure as a neutron absorber, while water is used as a neutron shielding material in the HI-TRAC.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-1 Rev. I

The shielding analyses were performed with MCNP5 [M.G] developed by Los Alamos National Laboratory (LANL). The source terms for the design basis fuels were calculated with the SAS2H [M.1] and ORIGEN-S [M.H] sequences from the SCALE 4.3 code system and were previously utilized in HI-STORM 100 FSAR [K.A]. These are principally the same codes that were used in Holtec's approved Storage and Transportation FSARs and SAR under separate docket numbers [K.A, K.B]. Detailed descriptions of the source term calculations and the MCNP models are presented in Sections 7.2 and 7.3, respectively.

The design basis fuel assemblies are a Babcock & Wilcox (B&W) 15x15 fuel assembly. While the fuel assembly type used at Indian Point Unit 3 is Westinghouse 15x 15, evaluations have shown that the B&W 15x 15 fuel assembly design is bounding when compared to other fuel assembly designs and classes. Therefore, the B&W 15x15 was used in this shielding evaluation. In addition, the B&W 15x15 fuel assembly design was used as the bounding design basis assembly in the HI-STORM 100 FSAR [K.A].

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-2 Rev. I

7.1 SHIELDING DESIGN 7.1.1 Design Features The principal design features of the STC with respect to radiation shielding consist of the fuel basket and STC body. The main shielding is provided by the canister body which includes steel and lead for gamma shielding and Metamic is included in the fuel basket for neutron absorption.

The drawings that describe the STC and show its dimensions can be found in Chapter 1.

The design basis fuel assembly is the B&W 15xl5 with source terms computed for this assembly at an initial enrichment of 3.6 wt % 235U. The design basis initial enrichment is a representative value, while the utilized source terms, shown in Table 7.1.1 and described in Section 7.2, are both representative and bounding of the expected discharged fuel.

7.1.2 Acceptance Criteria As discussed in Chapter 3, the acceptance criteria for the site boundary dose evaluation is 10 CFR 72.104 [A.C] for normal conditions and 10 CFR 72.106 [A.C] for accident conditions. The acceptance criteria from 10 CFR 72 were used rather than 10 CFR 100, since the 10 CFR 72 regulations are more restrictive. Both regulations are summarized below.

Normal condition requirements from 10 CFR 72.104.

During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area, must not exceed 25 mrem to the whole body, 75 mrem to the thyroid and 25 mrem to any other critical organ.

Accident condition requirements from 10 CFR 72.106.

Any individual located on or beyond the nearest boundary of the controlled area may not receive from any design basis accident the more limiting of a total effective dose equivalent of 5 rem, or the sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue (other than the lens of the eye) of 50 rem. The lens dose equivalent shall not exceed 15 rem and the shallow dose equivalent to skin or to any extremity shall not exceed 50 rem. The minimum distance from the spent fuel or high level radioactive waste handling and storage facilities to the nearest boundary of the controlled area shall be at least 100 meters.

In these calculations, 50 m is conservatively used as the site boundary.

In addition, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable (ALARA) shall be used per 10 CFR 20.1101 (b) [A.B].

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-3 Rev. I

Table 7.1.1 UTILIZED SOURCE TERMS (initial enrichment = 3.6 wt% U-235)

Loading Pattern Region Burnup (GWd/MTU) Cooling Time (years)

1. Representative Outer 8 cells 45 20 Inner 4 cells 55 10

2. Bounding Outer 8 cells 45 20 Inner 4 cells 55 5

3. Site Boundary* Outer 8 cells 55 5 Inner 4 cells 55 5

This loading is not permitted per the Technical Specifications, it is provided as a bounding scenario for the site boundary doses.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-4 Rev. 1

7.2 SOURCE SPECIFICATION 7.2.1 Source Term Selection These source terms were selected based on a survey of the current Indian Point Unit 3 spent nuclear fuel inventory. The selected source terms represents the most abundant/representative source term characteristics as well as a bounding source term scenario (Table 7.1.1). The initial enrichment is 3.6 wt%, which is conservative for high burnup fuel.

The source terms were applied in a regionalized loading scheme to the 12 fuel assembly locations available in the STC. The regionalized loading pattern was utilized to more easily be able to transfer hotter fuel in the spent nuclear fuel storage pool by taking advantage of self-shielding effects. The source terms with the higher cooling times are assigned to the eight outer fuel assembly locations in the STC, while the source terms with the lower cooling time are assigned to the four inner fuel assembly locations.

7.2.2 Principal Sources of Radiation The principal sources of radiation in the STC are the gamma and neutron radiation originating from various sources (e.g., decay of radioactive fission products, spontaneous fission). The neutron and gamma source terms were calculated with the SAS2H [M.I] and ORIGEN-S [M.H]

modules of the SCALE 4.3 code system using the 44-group library and have been previously utilized in the HI-STORM 100 FSAR [K.A]. In performing the SAS2H and ORIGEN-S calculations, a single full power cycle was used to achieve the desired burnup. All source term calculations were also performed assuming an infinite array of assemblies during irradiation. The design basis fuel assembly characteristics used in the computations as well as the modeling approach of the gamma and neutron sources are from the FSAR [K.A].

Table 7.2.1 provides the gamma source in MeV/s and photons/s as calculated with SAS2H and ORIGEN-S for the design basis burnup and cooling time combinations for the B&W 15x 15 assembly design. Per recommendations in NUREG- 1617 [C.F], only photons with energies in the range of 0.45 to 3.0 MeV are included in the shielding calculations. Photons with energies below 0.45 MeV are too weak to penetrate the steel of the cask, and photons with energies above 3.0 MeV are too few to contribute significantly to the external dose.

The primary source of activity in the non-fuel regions of an assembly arises from the activation of 59Co to 6°Co. Table 7.2.2 provides the 6°Co activity utilized in the shielding calculations in the non-fuel regions of the assemblies. The cobalt-59 impurity level was assumed to be 1 g/kg for all non-fuel hardware pieces, which is consistent with the HI-STORM 100 FSAR [K.A]. Further, the FSAR provides the scaling factors used in calculating the 60Co source along with a detailed description of the masses of the non-fuel regions.

Another source arises from (n,7) reactions in the material of the STC. This source of photons is properly accounted for in MCNP5 when a neutron calculation is performed in a coupled neutron-HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-5 Rev. I

gamma mode.

The neutron sources calculated for the design basis fuel assembly are listed in Table 7.2.3 in neutrons/s.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-6 Rev. I

TABLE 7.2.1 CALCULATED GAMMA SOURCE PER ASSEMBLY Lower Upper 45,000 MWd/MtU 45,000 MWd/MtU Energy Energy 10 Year Cooling 20 Year Cooling 3.6 wt% 235U 3.6 wt% 235U (MeV) (MeV) (MeV/s) (Photons/s) (MeV/s) (Photons/s) 0.45 0.7 1.33E+15 2.32E+15 9.64E+14 1.68E+15 0.7 1.0 1.64E+14 1.93E+14 2.69E+13 3.16E+13 1.0 1.5 6.90E+13 5.52E+13 2.71E+13 2.17E_+13 1.5 2.0 3.41E+12 1.95E+12 1.60E+12 9.16E+ 11 2.0 2.5 1.34E+11 5.97E+10 8.54E+09 3.80E+09 2.5 3.0 1.02E+10 3.72E+09 8.1OE+08 2.95E+08 Total 1.57E+15 2.57E+15 1.02E+15 1.73E+15 TABLE 7.2.1 (cont.)

CALCULATED GAMMA SOURCE PER ASSEMBLY Lower Upper 55,000 MWd/MtU 55,000 MWd/MtU 55,000 MWd/MtU Energy Energy 5 Year Cooling 10 Year Cooling 20 Year Cooling 3.6 wt% 215U 3.6 wt% 235 U 3.6 wt% 235 U

(MeV) (MeV) (MeV/s) (Photons/s) (MeV/s) (Photons/s) (MeV/s) (Photons/s) 0.45 0.7 2.62E+15 4.56E+15 1.63E+15 2.84E+15 1.17E+15 2.03E+15 0.7 1.0 9.57E+14 1.13E+15 2.18E+14 2.57E+14 3.43E+13 4.04E+13 1.0 1.5 2.14E+14 1.71E+14 8.99E+13 7.19E+13 3.49E+13 2.79E+13 1.5 2.0 1.22E+13 6.97E+12 4.35E+12 2.48E+12 2.03E+12 1.16E+12 2.0 2.5 7.38E+12 3.28E+12 1.47E+11 6.55E+10 9.84E+09 4.38E+09 2.5 3.0 3.43E+ 11 1.25E+ 11 1.25E+10 4.54E+09 1.29E+09 4.69E+08 Total 3.81E+15 5.87E+15 1.94E+15 3.17E+15 1.24E+15 2.10E+15 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-7 Rev. I

TABLE 7.2.2 CALCULATED 6"Co SOURCE 3 5U Location 45,000 MWd/MtU, 3.6 wt% 23 "U 55,000 MWd/MtU, 3.6 wt% 1 (Ci/1 g of Co-59) (Ci/1 g of Co-59) 10 Year 20 Year 5 Year 10 Year 20 Year Cooling Cooling Cooling Cooling Cooling Non-fuel hardware 4.57E+01 1.23E+01 1.05E+02 5.42E+01 1.45E+01 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-8 Rev. I

TABLE 7.2.3 CALCULATED NEUTRON SOURCE PER ASSEMBLY Lower Energy Upper Energy 45,000 MWd/MTU 45,000 MWd/MTU (MeV) (MeV) 10 Year Cooling 20 Year Cooling 3.6 wt% 235U 3.6 wt% 231U (Neutrons/s) (Neutrons/s) 1.01E-01 4.0E-01 1.37E+07 9.41E+06 4.OE-01 9.OE-01 7.OOE+07 4.81E+07 9.0E-01 1.4 6.41E+07 4.41E+07 1.4 1.85 4.73E+07 3.27E+07 1.85 3.0 8.38E+07 5.84E+07 3.0 6.43 7.58E+07 5.24E+07 6.43 20.0 6.70E+06 4.60E+06 Totals 3.61E+08 2.50E+08 TABLE 7.2.3 (cont.)

CALCULATED NEUTRON SOURCE PER ASSEMBLY Lower Energy Upper Energy 55,000 MWd/MTU 55,000 MWd/MTU 55,000 MWd/MTU (MeV) (MeV) 5 Year Cooling 10 Year Cooling 20 Year Cooling 3.6 wt% 235U 3.6 wt% 23

.U 3.6 wt% 131U (Neutrons/s) (Neutrons/s) (Neutrons/s) 1.OE-01 4.OE-01 3.59E+07 2.95E+07 2.02E+07 4.OE-01 9.OE-01 1.83E+08 1.51E+08 1.03E+08 9.OE-01 1.4 1.67E+08 1.38E+08 9.48E+07 1.4 1.85 1.22E+08 1.02E+08 6.99E+07 1.85 3.0 2.11E+08 1.79E+08 1.24E+08 3.0 6.43 1.95E+08 1.63E+08 1.12E+08 6.43 20.0 1.76E+07 1.45E+07 9.91E+06 Totals 9.32E+08 7.77E+08 5.34E+08 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-9 Rev. I

7.3 SHIELDING MODEL The shielding analysis of the STC was performed with MCNP5 [M.G]. MCNP is a Monte Carlo transport code that offers a full three-dimensional combinatorial geometry modeling capability including such complex surfaces as cones and tori. This means that no gross approximations were required to represent the STC in the shielding analysis. MCNP is the same code that has been used for shielding calculations of Holtec in previous dry storage and transportation systems licensing calculations.

The MCNP model of the HI-TRAC for normal conditions have the jackets filled with water, but does not credit the water for the hypothetical accident condition.

In the shielding analysis, the STC and the HI-TRAC are only partially filled with water leaving an air gap under the lids. The air gaps are needed to provide an expansion zone for the water and also allow the STC lid operations to occur unhindered by water.

The MCNP model contains some simplifications and assumptions. The simplifications include the omission of lifting or "operational" features of the STC and HI-TRAC, as these would not impact the dose rates significantly (leaving them out is conservative). The HI-TRAC was modeled with a lid consisting of a steel plate with a thickness of 3 in.

7.3.1 Configuration of Shielding and Source Chapter 1 provides the drawings that describe the STC and the HI-TRAC. These drawings were used to create the MCNP models used in the radiation transport calculations.

7.3.1.1 Shielding Configuration The normal conditions of shielding configuration for the STC and the STC placed inside the HI-TRAC is shown in Figures 7.3.1 through 7.3.4 for the 12 fuel assemblies. Steel and lead are considered as shielding materials for the STC shielding design. The fuel basket is modeled with steel and Metamic (see Chapter 1). It can be seen from Figures 7.3.2 and 7.3.4 that the STC and HI-TRAC are both partially filled with water and contain air gaps under the lids (to provide an expansion zone for the water).

The hypothetical accident shielding configuration for the HI-TRAC is the same as for normal conditions except that the water in the jackets is not credited.

7.3.1.2 Fuel and Source Configuration Design basis fuel assemblies are modeled in each of the twelve basket locations. Fuel assembly locations inside the STC are shown in Figures 7.3.1 and 7.3.3. The active fuel region is modeled as a homogenous zone. The bottom nozzle, plenum and top nozzle regions are also modeled as homogenous regions of steel using an effective density and water. The energy distribution of the HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-10 Rev. I

source term is used explicitly in the MCNP5 model. A different MCNP5 calculation is performed 60CO for each of the three source terms (fuel neutron, fuel gamma, and hardware 6"Co). The source in the hardware was assumed to be uniformly distributed over the appropriate regions.

The axial distributions of the fuel source term due to the burnup shape for the B&W 15x 15 PWR fuel assembly are taken from the HI-STORM FSAR [K.A].

7.3.2 Material Properties Composition and densities of the various materials used in the STC and the HI-TRAC shielding analyses are taken from the HI-STORM FSAR [K.A]. The lower end fitting material composition includes water to represent the partially water filled conditions inside of the STC.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-11 Rev. 1

FUEL BASKET STEEL FIGURE 7.3.1 SHIELDED TRANSFER CANISTER WITH 12 PWR BASKET - CROSS SECTIONAL VIEW AS MODELED IN MCNP HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-12 Rev. I

LEAD PLATE STEEL TOP LID AIR WATER FUEL

- ASSEMBLY STEEL BOTTOM

/

FIGURE 7.3.2 CROSS SECTION ELEVATION VIEW OF SHIELDED TRANSFER CANISTER AND 12 PWR BASKET AS MODELED IN MCNP HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-13 Rev. I

SHIELDED TRANSFER CANISTER'

_______ __ IH1TRAC FIGURE 7.3.3 SHIELDED TRANSFER CANISTER INSIDE HI-TRAC IOOD - CROSS SECTIONAL VIEW AS MODELED IN MCNP HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-14 Rev. I

STEEL TOP LID AIR LEAD PLATE

/

WATER FUEL

/ ASSEMBLY WATER JACKET HI-TRAC STEEL i BOTTOM HI-TRAC BOTTOM LID SHIELDED TRANSFER CANISTER FIGURE 7.3.4 CROSS SECTION ELEVATION VIEW OF THE SHIELDED TRANSFER CANISTER INSIDE HI-TRAC IOOD AS MODELED IN MCNP HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-15 Rev. I

7.4 SHIELDING AND ALARA EVALUATION 7.4.1 Methods The MCNP5 code [M.G] was used for all of the shielding analyses. MCNP is a continuous energy, three-dimensional, coupled neutron-photon-electron Monte Carlo transport code.

Continuous energy cross-section data is represented with sufficient energy points to permit linear-linear interpolation between these points. The individual cross section libraries used for each nuclide are those recommended by the MCNP manual. All of these data are based on either ENDF/B-V or ENDF/B-VI data. The large user community has extensively benchmarked MCNP against experimental data. Reference [S.A] is an example of the benchmarking that has been performed. MCNP is the same code that has been used as the shielding code in all of Holtec's dry storage and transportation analyses. Note also that the principal approach in the shielding analysis here is identical to the approach in licensing applications previously reviewed and approved by the USNRC.

The energy distribution of the source term, as described earlier, is used explicitly in the MCNP model. A different MCNP calculation is performed for each of the three source terms (neutron, decay gamma, and 60Co). The axial distribution of the fuel source term is based on the axial bumup distribution in HI-STORM FSAR [K.A]. This axial burnup distribution is representative of the fuel to be loaded. The 60Co source in the hardware is assumed as uniformly distributed over the appropriate regions.

The dose rates at the various locations were calculated with MCNP using a two-step process.

The first step was to calculate the dose rate for each dose location per starting particle for each neutron and gamma group in each basket region for each axial and radial dose location. The second step is to multiply the dose rate per starting particle for each energy group and location (i.e., tally output/quantity) by the source strength (i.e. particles/sec) in that group and sum the resulting dose rates for all groups in each dose location. The normalization of these results and calculation of the total dose rate from neutrons, fuel gammas or Co-60 gammas is performed with the following equation.

Tfna T ,j .. (Equation 7.4.1) j=l j=1 Fmi

where, Ttinal= Final normalized tally quantity (rem/h)

N = Number of groups.

M = Number of regions Tij = Tally quantity from particles originating in MCNP in group i and region j (particles/cm 2)

Fij = Fuel Assembly source strength in group i and region j (particles/sec)

Fmi = Source fraction used in MCNP (sdef card) for group i Note that dividing by Fmi (normalization) is necessary to account for the number of MCNP HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-16 Rev. I

particles that actually start in group i. Also note that Ti is already multiplied by a dose conversion factor in MCNP.

The standard deviations of the various results were statistically combined to determine the standard deviation of the total dose in each dose location. The estimated variance of the total dose rate, S 2tt,,,

is the sum of the estimated variances of the individual dose rates S2 i. The estimated total dose rate, estimated variance, and relative error [M.G] are derived according to Equations 7.4.2 through 7.4.5.

Ri-=F' T* (Equation 7.4.2) n

=*t (Equation 7.4.3) n TTotI = T (Equation 7.4.4)

I (Ri ix R 'rI 2Total otaT i=l Rota - ___________

TTotal Trotal Trotal (Equation 7.4.5)

where, i = tally component index n = total number of components TTotal = total estimated tally Ti = tally i component S2Total = total estimated variance S2 = variance of the i component Ri = relative error of the i component Rrotal = total estimated relative error Note that the two-step approach outlined above allows the accurate consideration of the neutron and gamma source spectrum, and the location of the individual assemblies, since the tallies are calculated in MCNP as a function of the starting energy group and the region of the assembly location, and then in the second step multiplied with the source strength in each group in each location. It is therefore equivalent to a one-step calculation where source terms are directly specified in the MCNP input files, except for the approximation that fuel is modeled as fresh U0 2 fuel in MCNP, with an upper bound enrichment.

Since MCNP is a statistical code, there is an uncertainty associated with the calculated values. In MCNP the uncertainty is expressed as the relative error that is defined as the standard deviation of the mean divided by the mean. Therefore, the standard deviation is represented as a HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-17 Rev. 1

percentage of the mean. The standard deviation of the result depends on the variance reduction parameters used in the analyses and the number of starting particles for each run. These parameters were chosen so that the relative error for the dose rates presented in this chapter was typically less than 2%.

7.4.2 Flux-to-Dose-Rate Conversion MCNP5 was used to calculate dose rates at the various desired locations. The point and ring detector tally, F5, as well as the surface tally, F2, were utilized in this calculation. MCNP5 calculates neutron or photon fluxes and these values can be converted into dose by the use of dose response functions. This is done internally in MCNP and the dose response functions are listed in the input file. The response functions used in these calculations were taken from ANSI/ANS 6.1.1-1977 [B.U].

7.4.3 External Radiation Levels -STC The calculated dose rate results for the STC are presented in Table 7.4.1 for a regionalized loading pattern. Total surface dose rates for regionalized loading using loading pattern 2 (Table 7.1.1) were found to be on the order of 2-3 rem/hr in the radial and top axial directions. The more representative loading pattern 1 (Table 7.1.1) reduces the surface dose rates by approximately 13%

in the radial direction and about 35% in the axial directions.

7.4.4 External Radiation Levels - HI-TRAC 100 D Transfer Cask The calculated dose rate results for the HI-TRAC 100D containing the STC are presented in Table 7.4.2 with regionalized loading pattern 2 (Table 7.1.1). The calculated HI-TRAC surface dose rates were found to be low, with the highest surface dose rate being 0.17 rem/hr in the top axial direction (top lid).

7.4.5 Dose Contribution to Site Boundary The dose contribution from the STC inside the HI-TRAC when transported between Unit 3 and Unit 2 was found to be 0.62 mrem under normal conditions. The distance from the HI-TRAC to the site boundary is conservatively estimated to be 50 m. This dose value is based on an estimated transportation time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . A uniform source term, loading pattern 3 (Table 7.1.1) was utilized for this calculation and the dose rate contribution from the radial and top HI-TRAC surfaces were considered, and provides a conservative bounding condition.

In the event an off-normal condition occurs such that the HI-TRAC would remain stationary between Unit 3 and Unit 2 for an extended period of time, the additional dose contributions to the site boundary should be considered. If the HI-TRAC remained stationary for an extended period of time, actual dose contributions should be estimated. In addition, temporary shielding may be used.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-18 Rev. 1

The dose contribution from the STC inside the HI-TRAC When transported between Unit 3 and Unit 2 was found to be 64.8 mrem under accident conditions. This dose value is based on an estimated "clean up" time of 30 days at 50 m away from the HI-TRAC surface consistent with the FSAR

[K.A]. The accident condition considers loss of water in the HI-TRAC jacket. The source term with 55 GWd/MTU burnup and 5 years cooling time was utilized for this calculation and the dose rate contribution from the radial and top HI-TRAC surfaces were considered.

The doses cited are the conservative equivalent to TEDE (without accounting for any self-shielding from the human body).

7.4.6 Effluent Dose Evaluation The consequence of an effluent dose release during normal and accident conditions has been analyzed using the calculational methodology in Chapter 7 of the HI-STORM FSAR, Revision 1

[K.A] which is based on the guidance in NUREG-l1536 [C.D] and ISG-5 [E.W] and NUREG/CR-6487 [C.P]. For normal conditions, a fraction of fuel rods are assumed to be breached in the STC causing a dose release over a period of time while for non-credible accident conditions all 12 fuel assemblies in the STC are assumed to be breached and causing an instantaneous release. Only gases are considered in the analysis, since the shielded transfer canister remains flooded during all operations, such that fines, volatiles and crud would remain entrapped within the water environment.

This is also consistent with the analysis of the fuel handling dose analysis in Chapter 14 of the Indian Point Unit 3 FSAR for the spent fuel pool.

The radiological inventory is based on the design basis assembly from the HI-STORM FSAR, Revision 1, which assumes a minimum cooling time of 5 years and an assembly average burnup of 70,000 MWD/MTU, which are bounding for the Indian Point Unit 3 fuel assemblies to be transferred. The radiological source term and other input parameters that are used in the analysis are provided in Table 7.4.4. These parameters are based on the following considerations:

" Dose Conversion Factors (DCF) from EPA Federal Guidance Report No. 11, Table 2.1 [T.J]

were used for the analysis.

The atmospheric dispersion factor is determined at 100 meters, based on Reg. Guide 1.145

[D.EE]. This distance is also consistent with 10 CFR 72. In addition, the resulting dose values, presented in Table 7.4.5, are so low that this utilized distance is' sufficient for the current case.

The cavity free volume is based on a 6" gap between the bottom of the shielded transfer canister lid and the enclosed water. This gap includes thermal expansion of the water after loading.

The duration of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months represents the transportation time between Unit 3 and Unit 2 under normal conditions, while 30 days represents accident conditions.

Doses from submersion in the plume are neglected because they are shown in [K.A] to be small compared to inhalation doses.

The release fractions for gases are based on ISG-5 [E.W], which specifies 0.30.

The leak rate testing performed on the STC seals verifies the leak rate to be less than or equal to lx 10-2 std-cm 3/s. This water-tight acceptance criterion is derived from ASTM HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-19 Rev. I

E1003-05, paragraph 11.3 [H.C]. However, it is conservatively assumed that the maximum possible leakage rate from the confinement vessel is 150% of the maximum leakage rate acceptance criteria.

Any ability of the HI-TRAC transfer cask to prevent the release of activity is neglected.

The activity released from the shielded transfer canister is assumed to be released at the maximum rate for one transfer between Unit 3 and Unit 2 during normal (8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months duration) and accident conditions (30 day, duration). Since no filtration or isolation of the release path is included in the model, this analysis supports a potential effluent release during all anticipated operations. The resulting doses are presented in Table 7.4.5. The doses are much smaller than from direct radiation (see Section 7.4.5). In addition, the resulting doses from the accident condition are significantly lower than the limits established in 10 CFR 72.106 (e.g., 5 rem TEDE, 50 rem TODE=CDE+DDE) and well below the doses rates calculated (for the site boundary and control room occupants) for the fuel handling accident presented in Chapter 14 of the IP3 UFSAR. Because the 10 CFR 72.106 limits are lower than those of 10 CFR 100.11, the part 100 limits are met by comparison.

7.4.7 Occupational Exposures for ALARA Consideration The overall personnel (person-rem) exposure from transferring one STC into the HI-TRAC is shown in Table 7.4.6. The values are estimates of exposures associated with different aspects of the transfer as described in Chapter 10. The occupational exposures received from transferring the STC in and out of the HI-TRAC are based on dose rates for regionalized loading with fuel bumups of 45 GWd/MTU and 20 years cooling time in the outer region and 55 GWd/MTU and 10 years cooling time in the inner region. Note that these dose rates are based on the usage of long-reach tools for the operators to prevent them from having to come in direct contact with the STC. Also, the dose received by the secondary personnel as shown in Table 7.4.6 is considered bounding for control room occupants.

7.4.8 Summary and Conclusions In summary, this calculation has demonstrated that the STC design promotes reasonable dose rates during the short period when the STC is moved from the SFP into the HI-TRAC. Further, it can be seen that the HI-TRAC surface dose rates are very low when containing the STC.

Based on the results shown in Table 7.4.6, the overall personnel (person-rem) dose from the operations necessary for moving the STC into the HI-TRAC is fairly low and is acceptable as part of the annual dose incurred at the plant and in accordance with ALARA. -

It can be concluded that the dose contribution to the site boundary from the STC is very small.

Further, the dose contributions for both normal and accident conditions from the STC are well below the regulatory limits.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-20 Rev. I

TABLE 7.4.1 TOTAL DOSE RATES AT VARIOUS DISTANCES AROUND THE SHIELDED TRANSFER CANISTER. REGIONALIZED LOADING PATTERN WITH FUEL BURNUP OF 45 GWD/MTU AND 20 YEARS COOLING TIME IN OUTER REGION AND 55 GWD/MTU FUEL BURNUP WITH 5 YEARS COOLING TIME IN INNER REGION.

6 Fuel °Co Dose Rate Gammas t Gammas Neutrons Total Location (mrem/hr) (mremhr) (mrem/hr) (mrem/hr)

Radial Surface of STC Surface 567.4 0.60 2573.8 3141.7 0.5 m away from surface 266.9 4.66 1044.4 1316.0 1 m away from surface 177.7 10.7 623.4 811.8 2 m away from surface 96.9 20.8 284.4 402.1 5 m away from surface 28.4 20.6 69.3 118.3 10 m away from surface 8.66 8.34 20.0 37.0 Top Lid of STC Surface 142.0 1477.1 414.7 2033.8 0.5 m away from surface 90.4 916.5 160.2 1167.1 1 m away from surface 45.8 399.7 69.0 514.5 10 m away from surface 1.12 6.87 2.71 10.7 Bottom Plate of STC Surface 347.4 6109.2 38.1 6494.7 1 m away from surface 177.5 3432.7 13.8 3624.0 10 m away from surface 2.97 48.7 1.93 53.6 t Dose rate contribution from the (n,p) reaction is included.

ItOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-21 Rev. I

TABLE 7.4.2 TOTAL DOSE RATES AT VARIOUS DISTANCES AROUND THE HI-TRAC 100D.

REGIONALIZED LOADING PATTERN WITH FUEL BURNUP OF 45 GWD/MTU AND 20 YEARS COOLING TIME IN OUTER REGION AND 55 GWD/MTU FUEL BURNUP WITH 10 YEARS COOLING TIME IN INNER REGION.

Dose Rate Fuel Gammas t 6

°Co Gammas T T Neutrons Total Location (mrem/hr) (mrem/hr) (mrem/hr) (mrem/hr)

Radial Surface of HIL-TRAC 100D Surface 0.25 0.001 0.79 1.03 1 m away from surface 0.11 0.003 0.28 0.39 10 m away from surface 0.008 0.002 0.02 0.03 50 m away from surface (site boundary) < 0.001 < 0.001 0.002 0.003 Top Lid of HI-TRAC 100D Surface 3.10 36.4 130.6 170.1 1 m away from surface 1.00 11.1 24.2 36.3 10 m away from surface 0.02 0.24 0.44 0.70 50 m away from surface (site boundary) 0.001 0.009 0.01 0.02 Bottom Lid of HI-TRAC 100D Surface (average) 0.15 4.58 4.85 9.58 1 m away from surface 0.28 10.2 3.52 14.0 10 m away from surface < 0.001 0.18 0.06 0.24 50 m away from surface < 0.001 0.006 0.003 0.009 Dose rate contribution from the (n,p) reaction is included.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-22 Rev. I

TABLE 7.4.3 TOTAL DOSE RATES AT VARIOUS DISTANCES AROUND THE HI-TRAC lOOD FOR ACCIDENT CONDITIONS. UNIFORM LOADING PATTERN WITH FUEL BURNUP OF 55 GWD/MTU AND 5 YEARS COOLING TIME.

Dose Rate Fuel 6°Co Gammas t Gammas Neutrons Total Location (mrem/hr) (mrem/hr) (mrem/hr) (mrem/hr)

Radial Surface of HI-TRAC 100D Surface 2.00 0.002 14.2 16.2 1 maway from 0.89 0.02 4.92 5.83 surface 50 m away 0.009 0.001 0.01 0.02 from surface Top Lid of HI-TRAC 100D Surface 15.9 128.8 240.1 384.8 1mawayfrom 6.10 45.5 44.8 96.1 surface 50 m away < 0.001 0.04 0.03 0.07 from surface Bottom Lid of HI-TRAC 100D Surface (vrace 1.28 18.7 9.42 29.4 (average) 1 maway from 1.89 36.0 6.91 44.8 surface 50 m away 0.002 0.02 0.008 0.03 from surface 0 Dose rate contribution from the (n,p) reaction is included.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT 141-2094289 7-23 Rev. I

TABLE 7.4.4 ANALYSIS INPUTS FOR EFFLUENT DOSE RELEASE ANALYSIS Parameter Value Fuel Assembly Inventory (Ci/Assembly)

-3H 3.68E+02 1291 3.31E-02 85Yr 5.86E+03 Fuel Rod Breakage Percentage 1% (normal conditions), 100% (large accident release)

Atmospheric Dispersion Factor 8.0 x10-3 (sec/mr3)

Leakage Rate (cm/sec) 1.5 x 10-2 Breathing Rate (m 3/sec) 3.30 x 10-4 Duration 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 30 days Free Volume (cm 3) 1.36 x 105 TABLE 7.4.5 DOSE FROM EFFLUENT RELEASE AT 100 METERS Effected Component Dose (mrem) Dose (mrem) Dose (mrem) 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months duration 30 day duration Instantaneous accident release Whole Body - TEDE 0.009 0.796 278.5 Max Organ (Thyroid) - TODE 0.065 5.83 2039.6 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-24 Rev. I

TABLE 7.4.6 PERSON-REM EXPOSURE FROM LOADING AND UNLOADING ONE SHIELDED TRANSFER CANISTER Number of Duration Estimated Person-Activity Personnel* (min) Dose Condition Rem Exposure Loading Transfer Canister into HI-TRAC (Unit 3)

Water filled STC in pool with Lid bolts installed on loaded 1 (primary) 1 meter of cask body 0.114 (primary) 20 exposed. Primary personnel STC 2 (secondary) 1 m distance. Secondary 0.088 (secondary) personnel 2 m distance.

Water filled STC (less 22.9 cm) in pool with 1 meter of Drain STC water for 1 (primary) 15 cask body exposed. Primary 0.085 (primary) movement 2 (secondary) personnel 1 m distance. 0.066 (secondary)

Secondary personnel 2 m distance.

Water filled STC (less 22.9 2 (primary) cm) in air with cask body 0.230 (primary)

Place STC into HI-TRAC 30 exposed. Primary personnel 3 2 (secondary) m distance. Secondary 0.080 (secondary) personnel 6 m distance.

Water filled STC (less 22.9 cm) in HI-TRAC with 0.5 Remaining STC lid bolts 2 (primary) 20 meter of cask body exposed. 0.0007 (primary) installed and tightened 2 (secondary) Primary personnel 0.5 m 0.0003 (secondary) distance. Secondary personnel 2 m distance.

Water filled STC (less 22.9 cm) in HI-TRAC with 0.5 Perform leak test of STC 1 (secondary) 15 meter of cask body exposed. 0.0002 (secondary) seal** Secondary personnel 0.5 m distance (5 min),and 1 m distance (10 min).

Water filled STC (less 22.9 cm) in HI-TRAC with lid and HI-TRAC top lid installed and 2 (primary) 30 0.5 meter of cask body 0.001 (primary) bolts are tightened 2 (secondary) exposed. Primary personnel 0.0004 (secondary) 0.5 m distance. Secondary personnel 2 m distance.

Unloading Transfer Canister out of HI-TRAC (Unit 2)

Water filled STC (less 22.9 cm) in HI-TRAC with lid and 1 (primary) 0.5 meter of cask body 0:0008 (primary) 4 (secondary) exposed. Primary personnel 3 0.0031 (secondary) m distance. Secondary personnel 6 m distance.

Water filled STC (less 22.9 2 (primary) cm) in HI-TRAC with lid and 0.001 (primary)

HI-TRAC lid is removed 30 0.5 meter of cask body 2 (secondary) exposed. Primary personnel .04 (secondary) 1 0.5 m distance. Secondary HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT Hl-2094289 7-25 Rev. I

Number of Duration Estimated Person-Personnel* (min) Dose Condition Rem Exposure personnel 2 m distance.

Water filled STC (less 22.9 cm) in HI-TRAC with 0.5 25 meter of cask body exposed. 0.0009 (primary)

Prepare STC lid 2 (secondary) Primary personnel 0.5 m 0.0003 (secondary) distance. Secondary personnel 2 m distance.

Water filled STC (less 22.9 STC raised & positioned in 2 (primary) cm) in air with cask body 0.192 (primary) 25 exposed. Primary personnel 3 SFP 2 (secondary) m distance. Secondary 0.067 (secondary) personnel 6 m distance.

Water filled STC in pool with 1 (primary) 1 meter of cask body 0.114 (primary)

STC lid bolts are removed 20 exposed. Primary personnel 2 (secondary) 1 m distance. Secondary 0.088 (secondary) nprqnnnpl 9 m riiqtnnrp 0.74 (primary)

Total Exposure, Person-Rem 0.39 (secondary)

Primary personnel consist of operators while secondary personnel include supervisors, qu, assurance staff, and health physicists.

- Leak tests will also be performed of the HI-TRAC top and bottom lid (see Chapter 10). Since the dose contribution/exposure is so small, these activities have been omitted from this table.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 7-26 Rev. I

CHAPTER 8: MATERIALS EVALUATION, ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 8.1 Introduction The suitability of the materials used in the manufacture of SSCs deployed to transfer the IP-3 fuel to the IP-2.pool is considered in this chapter. The NRC guidance documents, such as SFST-ISG-15 [E.O] and SFST-ISG- 11 [E.K] have been used in making the safety assessment that all materials used are suitable for their intended purpose. The materials used in the Shielded Transfer Canister (STC) design are the same as those used in HI-TRAC IOOD and similarly the materials used in the STC fuel basket are the same as those used in the MPC basket of HI-STORM 100 system. The composition of the materials of the STC and the HI-TRAC are provided in Chapter 7 of this report. The STC and the HI-TRAC will be lifted using ANSI N 14.6

[B.S] compliant lifting devices.

8.2 Materials Used Table 8.2.1 provides a listing of the materials whose stability during the transfer operations is necessary to ensure operational safety and reliability. The table also provides the information on the environment to which the material is subjected.

A brief description of the properties of the materials relevant to their suitability assessment is provided below.

i. Low Carbon Steel:

The carbon steel in the STC is ASME SA516 Grade 70,.SA515 Grade 70 or SA36. The material properties of SA516 Grade 70 and SA515 Grade 70 are shown in Tables 3.3.2 of the HI-STORM 100 FSAR [K.A]. The material properties of SA36 are shown in Table 3.3.6 of the HI-STORM 100 FSAR [K.A]. The material specified to be used in the STC has been used by Holtec in its dry storage casks for over a decade. All exposed steel surfaces will be coated with paint specifically selected for performance in the operating environments. Even without coating, no adverse reactions (other than nominal corrosion) have been identified. The material complies with ASME Section II requirement for added assurance of adequate long term performance.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-1 Rev. I

ii. Stainless Steel:

Alloy X is used within this licensing application to designate a group of stainless steel alloys.

Alloy X can be any one of the following alloys: Type 316, 316LN, 304, or 304LN. Alloy X shall be used for fabrication of STC basket and ASME SA 564-630 precipitation hardened for threaded connections. The material properties of Alloy X is provided in Appendix 1.A of CFSAR [K.A].The bolting material properties are shown in Table 3.3.4 of the CFSAR. Stainless steel has been extensively used in spent fuel racks supplied by Holtec and stored in spent fuel storage pools for over 20 years with both borated and unborated water with no adverse reactions.

The material complies with ASME Boiler and Pressure Vessel Code,Section II requirement for added assurance of adequate long term performance.

iii. Neutron Absorber: Metamic Metamic is an isotropic aluminum-based, macroscopically homogeneous neutron absorber material widely licensed for use for spent fuel reactivity control in dry and wet storage applications. Metallurgically, Metamic is a metal matrix composite (MMC) consisting of a matrix of aluminum reinforced with Type 1 ASTM C-750 boron carbide. Metamic is characterized by extremely fine aluminum and boron carbide powder.

Metamic has been subjected to an extensive array of tests sponsored by the Electric Power Research Institute (EPRI) that evaluated the functional performance of the material at elevated temperatures (up to 482°C /900'F) and radiation levels (1E+l 1 rads gamma). The results of the tests documented in an EPRI report [I.A] indicate that Metamic maintains its physical and neutron absorption properties with little variation in its properties from the unirradiated state.

The main conclusions provided in the above-referenced EPRI report are summarized below:

The metal matrix configuration produced by the powder metallurgy process with a complete absence of open porosity in Metamic ensures that its density is essentially equal to the theoretical density.

The physical and neutronic properties of Metamic are essentially unaltered under exposure to elevated temperatures (399°C - 482°C / 7500 F - 9000 F).

a No detectable change in the neutron attenuation characteristics under accelerated corrosion test conditions has been observed.

In addition, independent measurements of boron carbide particle distribution show extremely small particle-to-particle distancet and near-perfect homogeneity. I t Medium measured neighbor-to-neighbor distance is 10.08 microns according to the article, "Metamic Neutron Shielding", by K. Anderson, T. Haynes, and R. Kazmier, EPRI Boraflex Conference, November 19-20, 1998.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-2 Rev. I

Because of its performance in dry storage applications, all Holtec fuel baskets employ Metamic (authorized under USNRC Docket No. 72-1014). Metamic is also licensed for use in a Holtec transport cask certified by the USNRC (viz., Docket No. 71-9336).

Prior to utilizing Metamic in its fuel racks and multi-purpose canisters (MPC), Holtec International ran an extensive property verification program that is documented in a (proprietary) technical report [I.B]. The tests carried out by the company to evaluate Metamic's suitability for wet and dry applications corroborated the conclusions reached by EPRI [I.A].

Consistent with its role in reactivity control, all Metamic material procured for use in the Holtec fuel baskets will be qualified as important-to-safety (ITS) Category A item. ITS-A is equivalent to Safety Related in Part 50. ITS category A manufactured items, as required by Holtec's NRC-approved Quality Assurance program, must be produced to essentially preclude the potential of an error in the procurement of constituent materials and the manufacturing processes.

Accordingly, material and manufacturing control processes must be established to, eliminate the incidence of errors, and inspection steps must be implemented to serve as an independent set of barriers to ensure that all critical characteristics defined for the material by the cask designer are met in the manufactured product.

All manufacturing and in-process steps in the production of Metamic shall be carried out using written procedures. As required by the company's quality program, the material manufacturer's QA program and its implementation shall be subject to review and ongoing assessment, including audits and surveillances as set forth in the applicable Holtec QA procedures to ensure that all Metamic panels procured meet with the requirements appropriate for the quality genre of the fuel basket. Additional details pertaining to the qualification and production tests for Metamic are summarized inSection 8.1 of HI-STAR 60 SAR [K.E].

The Metamic neutron absorber material, proposed for use for the STC basket, is manufactured by the Holtec's Nanotec Materials Division (NMD) in Lakeland, Florida. Metamic has been subjected to rigorous tests by various organizations including Holtec International, and it has been approved by the USNRC for use in Holtec's dry storage systems (USNRC Docket 72-1014

[K.A]) as well as recent wet storage applications for Arkansas Nuclear One Units 1 and 2 (Dockets 50-313 and 50-368), Clinton (Docket 50-461), Diablo Canyon Units 1 and 2 (Dockets 50-275 and 50-323), St. Lucie Unit 2 (Docket 50-389), Turkey Point Unit 3 (Docket 50-250) and Cooper Nuclear Station (Docket 50-298).

iv. Seals : Elastomeric The STC's ability to retain its contents relies on an elastomeric seal in the top lid as shown in the licensing drawings in Section 1.5. The elastomeric seal chosen for STC must fulfill the principal requirements set down in the following:

- A reasonably uniform compression/decompression characteristic over the temperature range of interest (-40 0 C to 200 0C)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT 1-11-2094289 8-3 Rev. I

Adequate springback upon withdrawal of the compression load

Ability to withstand borated water environment

Excellent radiation resistance

Well adapted for joints required to withstand impulsive and impactive loads Seals used may be silicone, neoprene and similar elastomers. These seals have a useful service life term and should be replaced as necessary. These seals have performed satisfactorily in spent fuel pools and ambient environments. Seals are selected to meet temperature and pressure requirements. The seals shall be procured as an Important-to-Safety part. The interfacing seating surfaces of the elastomeric seals are stainless steel or clad with stainless steel to assure long-term sealing performance and to eliminate the potential for localized corrosion of the seal seating surfaces.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-4 Rev. I

Table 8.2.1 MATERIALS AND THEIR ENVIRONMENT Thermal- Stress* Radiation t Expected Duration in Component Material I.D. Hydraulic Environment Environment Each Transfer Operation, Environment hrs.

1. Shielded i. Low carbon steel Borated Water Low Elevated Normal 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months but could Transfer coated with @ < 1400 C be several days Canister Carboguard 890 ii. Metamic iii. Stainless steel iv. Elastomeric Seal (Table 3.1.1)

2. HI-TRAC i. Low carbon steel Air @ <1400 C Low Moderate Normal 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months but could coated with be several days Thermaline 450 for interior surface and with Carboguard 890 for exterior surface ii. Elastomeric Seal (Table 3.1.1)

Low means stress level below 1 / 6 th of yield strength; moderate means stress level below 1/2 of yield; high means greater than 1/2 yield.

t Low means <5 mR/hr; moderate means < 200 mR/hr; elevated means > 200 mR/hr.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-5 Rev. I

8.3 Degradation Mechanisms The potential degradation mechanisms considered in this evaluation are:

i. Chemical and galvanic reactions:

Carbon Steel:

In accordance with NRC Bulletin 96-04 [F.C], a review of the potential for chemical, galvanic, or other reactions among the materials of the STC System, its contents and the operating environments, which may produce adverse reactions, has been performed.

The STC utilizes low alloy and nickel alloy steels, carbon steels, materials. All of these materials have a long history of non-galvanic behavior within close proximity of each other. The internal and external steel surfaces of the STC are sandblasted and coated with a material which has been proven to preclude surface oxidation. The STC coating does not chemically react with borated water. The coating materials to be used on the carbon steel surfaces of the STC are provided in Table 8.2.1. This same coating is used on the HI-TRAC which has been placed in the spent fuel pools during many dry storage campaigns. Therefore, chemical or galvanic reactions involving the STC materials are highly unlikely and are not expected.

Stainless Steel:

The Shielded Transfer Canister fuel basket utilizes two primary materials (1) Metamic neutron absorber material and (2) Stainless Steel.

The capacity for being passivated is the strength of stainless steels. Steels with chromium content greater than 12% are easily passivated. The addition of nickel markedly facilitates passivation.

AISI Type 304 stainless steel contains a minimum of 18% chromium and 8% nickel.. The passive films of stainless steels range between 10 to 50 angstroms (0.04 to 0.2 microinches) thick.

(Peckner & Bernstein, pp 16-17 [J.D]). Of all types of stainless steels (i.e., austenitic, ferritic, martensitic, precipitation hardenable and twophase), "the austenitic stainless alloys are considered the most resistant to industrial atmospheres and acid media" [J.D]. The results of experimental evaluations of stainless steel corrosion confirm the validity of this statement.

Experimental corrosion data for AISI Type 304 and 316 stainless steels (Swedish Designations SIS- 14-2333 and SIS- 14-2343, respectively) are available from the Swedish Avesta Jernverk laboratory [J.D]. Corrosive media evaluated in these tests include 4% (40,000 ppm) and 20%

(200,000 ppm) boric acid solutions and water, all at boiling. Under the evaluated conditions, the tested steels are identified as "fully resistant", with corrosion rates of less than 0.1 mm per year.

An even more extensive set of experimental corrosion data is available from ASM International

[J.E]. For test conditions without rapid agitation, similar to conditions that would exist during STC fuel loading in a spent fuel pool, all austenitic stainless steels available for STC fuel basket HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-6 Rev. I

fabrication (i.e., AISI Types 304, 304L) are resistant to corrosion in boric acid and water. No structural effects from corrosion from treated and borated water environments are expected.

Various NRC Information Notices, Bulletins, Generic Letters, and Circulars [F.C] have been reviewed in an effort to gain additional industry experience on corrosion of stainless steels. It is recognized that stainless steels in borated water and treated water (demineralized water) environments are susceptible to loss of material due to pitting corrosion and cracking due to intergranular stress corrosion cracking (IGSCC) attack but these mechanisms depend greatly on the presence of halogens and oxygen or the presence of sulfates and oxygen coupled with high stress and high temperature. Spent fuel pool and treated water chemistry programs normally keep the concentrations of halogens and sulfates at very low levels for the very same reason of avoiding corrosion problems not only with spent fuel assemblies but with other systems such as those that are relied upon for the operation of the spent fuel pool. In addition, stringent controls on water conductivity, which is essentially a measure of impurities, further limits corrosion in treated and borated water environments. Borated and treated water are considered as having a negligible effect on the stainless steel.

Corrosion products cause "crud" deposits on fuel assemblies. Crud, which is stable in oxygenated solutions, is not likely to contain materials that can react with stainless steel in any appreciable amount. Crud may leave a slight film of rust on the interior surfaces of the STC during fuel loading and closure activities.

Stainless steels have been extensively used in spent fuel storage pools with both borated and unborated water with no adverse reactions or interactions with spent fuel.

Metamic:

Once passivated, alloys of aluminum are extremely resistant to chemical attack. Experimental corrosion data for aluminum and its alloys from ASM International [J.E] shows that these materials are resistant to corrosion in both boric acid solutions and water. With respect to solutions of boric acid, ASM states "aqueous solutions of 1 to 15% boric acid at 60'C (140-F) did not attack the aluminum alloys 1100, 3003 or 6061". The aluminum used in manufacturing Metamic is either alloy 1100 or alloy 6061. With respect to water, ASM states "the slight reaction that occurs initially ceases almost completely within a few days after development of a protective oxide film of equilibrium thickness. After this conditioning period, the amount of metal dissolved by the water becomes negligible." These statements from ASM describe the process of passivation.

Neutron absorber materials and stainless steel have been used in close proximity in wet storage for over 30 years. Many spent fuel pools at nuclear plants contain fuel racks, which are fabricated from neutron absorber materials and stainless steel materials, with similar geometries.

Not one case of chemical or galvanic degradation has been found in fuel racks built by Holtec.

This experience provides a sound basis to conclude that corrosion will not occur in these materials.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT BI-2094289 8-7 Rev. I

ii. Brittle fracture The STC basket is constructed from a series of stainless steels listed above. These stainless steel materials do not undergo a ductile-to-brittle transition in the minimum temperature range of the STC.

The STC enclosure vessel is constructed from carbon steel materials which is the same material used in Holtec's dry storage transfer cask HI-TRAC. The HI-TRAC brittle fracture is analyzed in Subsection 3.1.2 of the HI-STORM 100 FSAR [K.A].

iii. Fatigue The STC is designed for repeated normal condition handling operations with high factor of safety, particularly for the lifting trunnions, to assure structural integrity. The resulting cyclic loading produces stresses that are well below the endurance limit of the trunnion material, and therefore, will not lead to a fatigue failure. All other off-normal or postulated accident conditions are infrequent occurrences that do not contribute significantly to fatigue. In addition, the STC

,utilizes materials that are not susceptible to brittle fracture during the lowest temperature permitted for loading.

The STC fuel basket is subject to cyclic temperature fluctuations. These fluctuations result in small changes of thermal expansions and pressures inside the STC. The loads resulting from these changes are small and do not significantly contribute to the "usage factor" of the STC.

As described in Chapter 6, the STC trunnions are designed to ANSI N 14.6 and procedures will implement operating and inspections requirements from ANSI N14.6.

iv. Stress corrosion/cracking Temperature distribution results obtained from this highly conservative thermal model show that the maximum local STC basket temperature level is below the recommended limits for structural materials in terms of susceptibility to stress, corrosion and creep-induced degradation.

Furthermore, stresses induced due to imposed temperature gradients are within Code limits (See Structural Evaluation Chapter 6).

v. Loss of neutron capture capability Unlike silicone polymer type neutron absorber material such as Boraflex, which has a-history of degradation under radiation in wet storage use, Metamic neutron absorber is a metal matrix composite (MMC) consisting of a matrix of aluminum reinforced with Type 1 ASTM C-750 boron carbide. Metamic is characterized by extremely fine aluminum (325 mesh or smaller) and boron carbide (B 4 C) powder. Typically, the average B4 C particle size is between 10 and 40 microns. The high performance and reliability of Metamic derives from the fineness of the B 4 C particle size and uniformity of its distribution, which is solidified into a metal matrix composite HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-8 Rev. 1

structure by the powder metallurgy process. This yields excellent homogeneity and a porosity-free material. As stated earlier Metamic has been extensively studied and characterized by multiple independent organizations over many years including EPRI and Holtec. None of the test results have shown any sign of degradation or loss of neutron capture capability in the Metamic.

USNRC has approved the use of Metamic for a number of plants mentioned in the previous sections.

vi. Generation of flammable gases and risk of combustion Because Metamic is a solid material there is no capillary path through which spent fuel pool water can penetrate Metamic panels and chemically react with aluminum in the interior of the material to generate hydrogen. Any chemical reaction of the outer surfaces of the Metamic neutron absorber panels with water to produce hydrogen occurs rapidly and reduces to an insignificant amount in a short period of time. Since the STC lid is bolted and there are no welding and cutting operations involved, there is no concern for combustion.

vii. Swelling of neutron absorber Because Metamic is a porosity-free material, there is no capillary path through which spent fuel pool water can penetrate Metamic panels and chemically react with aluminum in the interior of the material to generate hydrogen. Thus, the potential of swelling and generation of significant quantities of hydrogen is eliminated.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-9 Rev. 1

8.4 Acceptance Tests In this section the inspections and acceptance tests to be performed on the STC prior to its use are summarized. These inspections and tests provide adequate assurance that the STC has been fabricated, assembled and accepted for use and loading under the conditions specified in this report.

Inspections and acceptance tests for the HI-TRAC 1 OD are contained in the HI-STORM 100 FSAR [K.A].

8.4.1 Visual Inspections and Measurements The STC shall be assembled in accordance with the licensing drawings supplied in Section 1.5.

The drawings provide nominal dimensions that define the limits on the dimensions used in licensing basis analysis. Fabrication drawings will provide any additional dimensional tolerances necessary to ensure component fit-up. Visual inspections and measurements shall be made and controls shall be exercised to ensure that the packaging conforms to the dimensions and tolerances specified on the fabrication drawing. These dimensions are subject to independent confirmation and documentation in accordance with the Holtec QA program approved in NRC Docket No. 71-0784.

The following shall be verified as part of visual inspections and measurements:

Visual inspections and measurements shall be made to ensure that the effectiveness is not significantly reduced. Any important-to-safety component found to be under the specified minimum thickness shall be repaired or replaced as required.

" Visual inspections shall be made to verify that neutron absorber panels are present as required by the basket design.

The packaging shall be inspected for proper cleanliness and preparation for use in accordance with written and approved procedures.

The visual inspection and measurement results for the STC shall become part of the quality documentation package.

8.4.2 Weld Examination The examination of STC welds shall be performed in accordance with the drawings in Section 1.5 and applicable codes and standards. Weld examinations and repairs shall be performed in accordance with applicable codes and standards. All weld inspections shall be performed in accordance with written and approved procedures by personnel qualified in accordance with SNT-TC-1A [H.B]. All required inspections, examinations, and tests shall become part of the final quality documentation package.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-10 Rev. I

8.4.3 Structural and Pressure Tests The STC shall be tested by a specified combination of methods as required by Section III, Subsection ND of the ASME B&PV Code, to verify that it is free of cracks, pinholes, uncontrolled voids or other defects that could significantly reduce the effectiveness during its service life.

8.4.4 STC Leakage Test STC leakage testing shall be performed per written and approved procedures and in accordance with the requirements of ANSI N14.5 [B.T]. Criteria and basis for leak rate and leakage testing are discussed in Chapter 7 of this report.

In case of an unsatisfactory leakage rate, weld repair, seal surface cleaning/repair, or seal change and retest shall be performed until the test acceptance criterion is satisfied.

Leakage testing results shall become part of the quality documentation package.

8.4.5 Component and Material Tests The STC closure seal is an elastomeric seal and conservatively specified to provide a high degree of assurance of sealing function under normal and accident conditions. Seal tests under the most severe service conditions including performance at pressure under high and low temperatures will not challenge the capabilities of these seals and thus are not required.

The majority of the STC materials are ferritic steels. ASME Code Section III requires that certain materials be tested in order to assure that these materials are not subject to brittle fracture failures. Test results shall become part of the final quality documentation package.

ftOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-11 Rev. 1

8.5 Maintenance Program An ongoing maintenance program is defined and incorporated in the Operations & Maintenance (O&M) Manual, which will be prepared and issued prior to the delivery and first use. This document shall delineate the detailed inspections, tests, and parts replacement necessary to ensure continued radiological safety, proper handling, and continued performance of the STC in accordance with the design requirements and criteria contained in this report.

There are no active components or systems required to assure the continued performance of the safety functions. As a result, only minimal maintenance will be required over its lifetime, and this maintenance would primarily result from weathering effects and pre- and post-usage requirements. Typical of such maintenance would be the removal of scratches, dents, etc. from accessible external surfaces to eliminate locations for potential contaminant hideout; seal replacement; and re-coating surfaces. Such maintenance requires methods and procedures no more demanding than those routinely used at power plants.

A maintenance program schedule is provided in Table 8.5.1.

8.5.1 Structural and Pressure Tests Periodic structural or pressure tests on the cask following 'the initial acceptance tests are not required to verify continuing performance.

8.5.2 STC Leakage Test The elastomeric seals on the STC and HI-TRAC lid shall be replaced as defined in Table 8.5.1.

The STC seal will be tested prior to each fuel transfer.

8.5.3 Component and Material Tests 8.5.3.1 Surfaces Accessible external surfaces shall be visually inspected prior to each fuel loading for surface (superficial) and component damage including surface denting, surface penetrations, weld cracking, chipped or missing coatings, etc. Where necessary, any damage shall be restored per the instructions in Holtec supplied O&M manual. Damage to components shall be evaluated for impact on safety and components shall be repaired or replaced accordingly. Wear and tear from normal use will not impact safety. Repairs or replacement in accordance with written and approved procedures, as set down in the O&M manual, shall be required if unacceptable conditions are identified.

Prior to installation or replacement of a seal, the cask sealing surface shall be cleaned and visually inspected for scratches, pitting or presence of an unacceptable surface finish. The HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-12 Rev. I

affected surface areas shall be restored as necessary in accordance with written and approved procedures.

8.5.3.2 Bolts Closure bolting shall be visually inspected for damage such as excessive wear, galling, or indentations on the threaded surfaces prior to installation. The severity of thread damage shall be evaluated as set forth in the STC O&M manual. Damaged bolting and/or fasteners shall be replaced accordingly. Closure lid bolting shall be replaced after every (no more than) 240 bolting cycles. One bolting cycle is the complete sequence of torquing and removal of bolts.

8.5.3.3 Lifting Devices Lifting devices shall be inspected prior to each fuel loading. The accessible parts and the local areas surrounding their attachments shall then be visually examined to verify no deformation, distortion, or cracking has occurred. Any evidence of deformation (other than minor localized surface deformation due to contact pressure), distortion or cracking of the lifting device or adjacent cask areas shall require repair. All special lifting devices shall be maintained and inspected in accordance with ANSI N 14.6.

8.5.3.4 Closure Seals The closure seals are shipped from the factory pre-inspected and carefully packaged. Prior to each use, inspect seals and sealing surfaces for conditions the that effect sealing capability, replace seals and/or repair surfaces as required.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-13 Rev. I

Table 8.5.1 Maintenance and Inspection Program Schedule Task Frequency External surface (accessible) visual Prior to each fuel transfer inspection Bolting visual inspection Prior to each fuel transfer Lifting devices visual inspection Prior to each fuel transfer Leakage Test of STC and HI-TRAC seals. Following each fuel loading, and prior to transfer. Acceptance criteria in accordance with requirements in Chapter 7 STC and the HI-TRAC lid seal As necessary based on inspection results or replacement failure to seal.

Closure bolt replacement Every 240 bolting cycles STC/HI-TRAC seal inspection Prior to each fuel transfer HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 8-14 Rev. 1

CHAPTER 9: ECONOMIC AND ENVIRONMENTAL CONSIDERATIONS Referenced as USNRC GL 78-11, OT Position, [T.A], Article V specifies environmental and economic considerations as essential elements of spent fuel storage and handling licensing amendment applications. This chapter provides justification for selecting the option for fuel transfer from IP-3 to IP-2 as the appropriate means to fulfill the directive of the GL 78-11. The fuel transfer process meets the 10 CFR 51.22(c)(9) criteria for categorical exclusion from environmental evaluation. The requested change will have no impact on the environment. The proposed change does not involve a significant hazards consideration. The proposed change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed change does not involve a significant increase in individual or cumulative occupational radiation exposure.

9.1 Environmental Consideration The environmental impact of plant operations were assessed in the Indian Point Final Environmental Statements for IP-2 [T.B] and IP-3 [T.C]. Each document was issued prior to the commercial operation of the respective units. This assessment included spent fuel handling and storage..

The proposed amendment would not alter the type or amount of nuclear fuel that can be received, used, and possessed at the sites. Limitations on the type and amount of fuel that can be stored in the IP-2 spent fuel pool and the manner in which it may be stored and handled would also not be changed. Only the IP-3 fuel cooled for at least 5 years in the spent fuel pool after being discharged from the reactor would be permitted for transfer.

It has been determined that the proposed operating license amendment, described in this report, will not:

" result in any impact on the plant's environment, including the water, land, or air;

consume more than a minute portion of the world's supply of raw material of stainless steel, carbon steel, lead, boron carbide, or aluminum;

" produce any new radioactive waste;

" produce any harmful gaseous, particulate, or liquid emissions;

require any new and hazardous activities by manufacturing or plant personnel; o result in significant increase in occupational radiation exposure; or

result in significant increase in radiation exposure to the public.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 9-1 Rev. 1

Each of the above conclusions is explained below:

9.1.1 Environmental Impacts The proposed change does not impact the current spent fuel pool or the spent fuel pool cooling and cleanup systems. The current licensed spent fuel pool capacity and supporting analysis for both IP-2 and IP-3 would not be changed. The capacity would be controlled by transferring fuel from the IP-3 SFP to the IP-2 SFP. Fuel will be removed as necessary from the IP-2 SFP and placed into dry fuel storage as has been done in the recent past. The fuel transfer process would not adversely affect the previously evaluated impact on the water, land or air usage at the site.

Conservative estimates of consumption of raw materials for the manufacturing of the Shielded Transfer Canister (STC) are as follows:

Carbon and Stainless Steel: <25,000 lbs Lead: <25,000 lbs Aluminum: <2,000 lbs Boron Carbide: <1,000 lbs Stainless Steel Weld Wire and Flux: <1,000 lbs Argon Gas (for welding): < 10,000 SCF The annual worldwide consumption of each of the above materials is at least 500,000 times as much as the, quantities needed for the proposed activity. Therefore, it is concluded that the proposed activity will have a negligible effect on the worldwide availability of the above-mentioned raw materials.

The STC is merely a temporary storage device; that produces no radioactive waste. After decades of use, a slight activation of the metals adjacent to the spent fuel is expected to occur.

However, the extent of activation is expected to be small enough to classify the STC as Low Specific Activity (LSA) at its retirement (at the end of the plant's operating life).

There is no known mechanism for the new STC to generate hazardous gases. The postulated creditable human and natural accident events for the transfer process have be evaluated in Chapter 3 and are bounded by the existing accidents conditions documented in the IP-2 and IP-3 FSARs [T.E and T.F]. Both the STC and the HI-TRAC 100D Transfer Cask include bolted closures with elastomeric seals which create redundancy in controlling particulate or liquid release of radioactive materials to the environment.

Manufacture of STC and supporting ancillary equipment and the installation of supporting facility modifications are routine and commonplace activities. Thousands of similar fuel storage canisters, racks, and shipping casks have been manufactured and are in use in the world's nuclear reactors. There is no scientific evidence (or assertion by any group) that the manufacture or use of fuel canister systems entails any human risk factors.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT 2094289 9-2 Rev. 1

9.1.2 Occupational Radiation Exposure A 10 CFR Part 50 licensed STC would be used to transfer up to twelve (12) PWR fuel assemblies at a time. The ALARA time, distance, and shielding principles would be used to limit occupational exposure. The fuel is loaded into the STC underwater. The STC would be moved remotely from the spent fuel pool to the HI-TRAC 1OOD Transfer Cask using the overhead crane. Controls would be in effect to reduce the possible spread of radioactive contamination. The HI-TRAC provides effective shielding and physical protection of the fuel during transfer between the IP-3 and IP-2 Fuel Storage Buildings (FSB). The occupational radiation exposure from the STC and HI-TRAC has been evaluated in Chapter 7 and are considered ALARA pursuant with 10 CFR Part 20.

The occupational radiation exposure for the fuel transfer operation is estimated to be less than 1.2 person-rem per transfer of 12 spent fuel assemblies. This small increase in radiation dose would not affect the ability to maintain individual occupational doses within the limits of 10 CFR Part 20 and is as low as is reasonably achievable (ALARA). The site plant radiation protection program is implemented by procedures, specifically the ALARA Program procedure

[T.I] that is in compliance with the guidelines of Regulatory Guide 8.8 [D.E] to preclude any significant occupational radiation exposure.

Based on the plant operations, the proposed fuel transfer between IP-3 and IP-2 should add only a small fraction of the total annual occupational radiation dose at the facility. The total twenty-four (24) month occupation dose for 2007 and 2008 at the site was approximately 137 person-rems. The total collective dose for the typical transfer of 96 spent fuel assemblies in one operation cycle would be 9.6 person-rem. This is a small percentage of the total average occupational radiation dose for the site and would not result in any significant increase to the occupational radiation doses received by plant workers.

9.1.3 Public Radiation Exposure The STC is placed in the HI-TRAC IOOD Transfer Cask for the transfer between the IP-3 and IP-2 FSBs. The radiation dose rates on the side, top and bottom of the HI-TRAC for the surfaces and at increasing distances are reported in Chapter 7. The impact on the site boundary dose would be negligible considering the low radiation dose rates at the assumed site boundary and the short period of time required for the transfer. The transfer haul path will be inside the plant protected area which is well within the site boundary. The typical time the loaded STC and HI-TRAC would be outside of either FSBs (on the haul path) is eight (8) hours. Assuming the transfer of 96 spent fuel assemblies or 8 transfers in one year, the estimated annual radiological dose commitments to a maximally exposed individual at the site boundary (conservatively assumed 50 meters from the HI-TRAC) due to the fuel transfer would be approximately 5.0 mrem. This estimated total annual dose commitment is within the limitation of the IP-2 and IP-3 Technical Specification [T.G and T.H], which are based on offsite dose requirements of 10 CFR Parts 20, 50, and 40 CFR 190.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 9-3 Rev. I

9.2 Economic Considerations In the year 2000 after evaluating alternatives for spent fuel storage, Entergy concluded that implementing an Independent Spent Fuel Storage Installation (ISFSI) pursuant to 10 CFR Part 72 using a general license was the best option to ensure continued reliable operation of IP-2. The Holtec HI-STORM 100 Cask System was selected and implemented for IP-2 in the year 2007.

The capacity of the IP-3 SFP is now reaching its limits. The IP-3 reactor holds 193 fuel assemblies. The plant operates on a twenty-four (24) month operating cycle. At the end of each cycle approximately 90 to 96 fuel assemblies are replaced with fresh fuel during the refueling outage. The IP-3 spent fuel pool has a storage capacity of 1,345 fuel assemblies. After the core reload from the 3R15 (spring 2009) refueling outage, the IP-3 spent fuel pool only has 103 available cells for future storage; thus full core offload capability has been lost. To provide full core off load capability for the next refueling outage 3R16 (Spring 2011) approximately 96 fuel assembles must be removed from the spent fuel pool prior to the outage. Going forward, to maintain full core off load capability for every refueling outage, approximately 96 fuel assemblies must be removed from the spent fuel pool each operating cycle prior to the refueling outage.

Entergy has evaluated alternatives for storing the excess IP-3 fuel. The following subsections describe the economic considerations of certain alternatives for spent fuel storage and are evaluated in two sections; first, the spent fuel storage options available and second, the implementation of dry fuel storage options for IP-3.

9.2.1 Spent Fuel Storage Options Evaluated 9.2.1.1 High Density Spent Fuel Pool Racks The original low density racks in the IP-3 spent fuel pool have been replaced with high density racks supplied by Holtec International. Additional expansion to increase storage capacity is limited due to available space.

The only open space in the spent fuel pool is set aside for the cask handling area. A temporary rack with a capacity of approximately 64 cells could be installed in the cask handling area of the spent fuel pool. However, there are interferences arising from fuel handling tools which would need to be relocated. The rack would eventually have to be removed to allow for cask handling for other dry fuel storage options. Full core off load capability would not be restored due to the limited capacity of the temporary rack. Another option would need to be identified prior to the 2013 refueling outage.

9.2.1.2 Fuel Rod Consolidation Spent fuel rod consolidation represents another potential option for expanding the spent fuel storage capacity. Spent fuel consolidation is the process by which spent fuel assemblies are disassembled in the spent fuel pool, and the fuel rods and other components of the fuel bundle are then repackaged into appropriate containers for storage in the pool. The fuel rods are HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 9-4 Rev. 1

removed individually, or in groups, from fuel bundles, depending on the equipment used, and are placed in a closely packed array into a fuel rod container with dimensions similar to that of a fuel bundle. These containers should fit in the existing fuel storage racks. It has been shown through several demonstration projects with PWR fuel that a compaction ratio of fuel rods of 2:1 is achievable.

At this time, proven technology to consolidate fuel does exist. The time involved in performing all of the initial functional tests, cold demonstrations, and hot demonstrations to achieve a reliable, workable system would exceed the time that Indian Point has available prior to needing additional storage capacity. The project costs exceed the costs for dry cask storage solutions including the fuel transfer. Therefore, this option is not viable for expanding the spent fuel storage capacity.

9.2.1.3 New Spent Fuel Pool Storage Building Extension of the storage pool would entail extensive modifications. The FSB currently could not support a larger pool, so an addition to the building or a new building would be required. A new safety related cooling system would have to be installed. These two challenges make the option of extending the storage pool cost and time prohibitive.

9.2.1.4 Dry Cask Storage Dry Cask Storage is a method of spent fuel storage that removes the spent fuel from the pool and stores it in metal canisters within a concrete overpack. This method permanently removes the fuel from the pool enabling continued operation without modifications to the pool or its associated systems. The casks are stored on a concrete storage pad which is specifically designed for the casks, also known as the ISFSI. The casks can store 24 to 32 PWR assemblies each depending on the vendor. This is a modular storage option so the casks can be purchased as needed.

Entergy facilities, FitzPatrick, Arkansas Nuclear One, River Bend Station, Grand Gulf and Indian Point Energy Center have conducted several studies between 1990 and 2008 on the spent fuel storage issue. These studies explored options similar to those presented above and addressed additional options such as transshipment and construction of new storage pools. In each case, dry fuel storage was selected as the best strategic option.

9.2.2 Options to Implement Dry Cask Storage at IP-3 Entergy chose the Holtec HI-STORM 100 Cask System as the dry cask storage (DCS) technology for use at the ISFSI. The HI-STORM 100 Cask System, like all contemporaneous DCS technologies, accommodates many more fuel assemblies than the 40-ton shipping cask assumed in original plant design and licensing. The 32-assembly multi-purpose canister (MPC) was chosen as the best, most cost-effective option for use at both units. Entergy understood at that time that the cask handling cranes in Units 2 and 3 were of insufficient capacity to support DCS operations and a crane upgrade would need to be addressed as part of the DCS projects for each plant.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT 1-11-2094289 9-5 Rev. I

The HI-STORM DCS operations involve the handling of heavy loads inside the power plants in the same areas designated for the shipping cask to facilitate movement of spent fuel from the spent fuel pools to the ISFSI. The DCS canister and transfer cask assemblage weighs about 100 tons when loaded and must be moved within the FSB as follows:

From a staging area into the spent fuel pool for fuel loading

From the spent fuel pool to the truck bay floor for canister closure operations after fuel loading

From the truck bay floor closure area to the truck bay floor where the transfer cask is stacked atop the storage overpack and the canister inside is transferred to the storage overpack

The transfer cask is removed from atop the overpack at the stack-up location and moved back to the other end of the truck bay floor.

" The loaded overpack is moved to the ISFSI using a suitably designed vertical cask transporter.

9.2.2.1 Crane Solutions Neither the IP-2 nor the IP-3 FSB structure could withstand the significant increase in loads that would result from increasing the capacity of the existing overhead bridge-and-trolley cask handling cranes from 40 tons to 100 tons. The under-capacity of the IP-2 and IP-3 cask handling cranes was addressed in different ways because the two plants are not identical with respect to the immediate site topography adjacent to the FSBs.

At IP-2, a floor-mounted gantry Crane was designed, fabricated, licensed, and installed to facilitate the handling of the approximate 100-ton lifted load comprised of the transfer cask and the MPC, which is filled with 32 fuel assemblies and spent fuel pool water. At IP-3, a similar crane upgrade was determined not to be feasible for the following primary reasons:

The distance between the top of the spent fuel pool pit wall and the truck bay is approximately 23 feet more at IP-3 than at IP-2. This increased distance significantly increases the amplification of seismic loads on the crane. Due to the limited space in the truck bay area, the significant required increase in crane structural member sizes may not be achievable. This makes the feasibility and cost of the upgrade, even if physically possible, imbalanced with the desired outcome.

o There are numerous plant equipment interferences that would require significant design and construction effort to re-locate.

With the constraint of the 40-ton FSB crane, the implementation of dry cask storage is limited to performing a fuel transfer from IP-3 to IP-2, then conducting dry cask storage from IP-2 or designing a light weight dry cask storage system. These options are evaluated in the next three subsections.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 9-6 Rev. 1

9.2.2.2 Dry Fuel Transfer with a licensed 10CFR71 Shipping Cask With consideration for the 40-ton lifting capacity of the Unit 3 FSB crane, the choices for a licensed 10 CFR Part 71 shipping cask are limited. This limitation would restrict the size of the cask to one that typically holds only one (1) PWR fuel assembly. One such approved Part 71 shipping cask is the NAC-LWT, Certificate of Compliance 71-9225, which could be used for the Indian Point inter-unit fuel transfer. The LWT cask has initial uranium enrichment limitations on the approved contents that do not support the total population of the IP-3 spent fuel inventory.

Using the LWT cask would require twelve (12) times the personnel and time resources resulting in approximately five (5) times the total occupation radiation exposure to complete each fuel transfer campaign. If multiple LWT casks are used, the overall transfer time could be reduced by forty percent, but the resources needed would increase and the occupational radiation exposure would be similar. Based on the duration and increased occupation radiation exposure this option is considered not practical or cost effective as a long time solution. Also using a Part 71 cask does not support the schedule for the first fuel transfer campaign prior to the 3R16 (Spring 2011) refueling outage.

9.2.2.3 Fuel Storage with a licensed 10CFR72 Light Weight Dry Storage System A light weight dry cask storage system could potentially be designed with a capacity of eight (8) to twelve (12) fuel assemblies. A new 40-ton transfer cask, similar to the HI-TRAC 100D Transfer Cask, would be required. The MPC would be designed similar to the MPC-24, but would have a maximum capacity of twelve (12) fuel assemblies. A special design overpack or insert for the existing HI-STORM overpack would be required. If the same overpack is used, three (3) times the ISFSI Pad storage space would be required for the same amount of spent fuel.

If a smaller overpack is designed, the required ISFSI Pad space would be approximately two (2) times that of the current HI-STORM 100 Cask System.

The use of this light weight dry cask storage system would require three (3) times the personnel hours and time resources resulting in approximately three (3) times the total occupation radiation exposure to complete each fuel storage campaign. To design this light weight dry cask storage system would take approximately one year to two years for NRC licensing reviews and rulemaking to add the new system to. the 10 CFR 72.214 list of approved spent fuel storage casks.

Based on the increased amount of MPCs per each dry storage campaign, and resulting occupational radiation exposure and storage space required the option of a light weight dry cask storage system is considered not cost effective. Also the Part 72 licensing effort does not support the schedule for the first fuel transfer campaign prior to the 3R16 (Spring 2011) refueling outage.

9.2.2.4 Wet Fuel Transfer with a licensed 10CFR50 Shielded Transfer Canister This option is described in Chapter 1 and uses a STC for transferring up to twelve (12) fuel assemblies between the IP-3 and IP-2 SFPs using the existing HI-TRAC 100D Transfer Cask and vertical cask transporter for the transfer*operation. The design and license acceptance criteria are documented in Chapter 3 and present this option as a 10 CFR Part 50 license amendment. The HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT Hl-2094289 9-7 Rev. I

STC capacity of twelve (12) fuel assemblies is optimized for occupational radiation exposure and amount of fuel assemblies transferred each evolution. Essentially this is a mobile spent fuel pool that provides equivalent or better protection of the spent fuel than in the existing spent fuel pool. The technical analyses supporting the license amendment are very similar to that of a spent fuel rack expansion license amendment. Credit for previously reviewed and approved HI-TRAC lOOD Transfer Cask analysis, under 10CFR72, is being used when applicable to support the on-site transfer operations. Based on reasonable occupational radiation exposure, cost effective movement of fuel, and subsequent placement into dry cask storage using a proven technology this option is considered to be the best strategic solution for the IP-3 spent fuel storage issue.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 .9-8 Rev. I

CHAPTER 10: OPERATING PROCEDURES 10.0 Introduction This chapter outlines the loading, unloading, and recovery procedures for the Shielded Transfer Canister (STC) in support of fuel transfer operations. The procedures provided in this chapter are prescriptive to the extent that they provide the basis and general guidance for preparing detailed, written, site-specific, loading, handling, and unloading procedures. Section 10.1 provides the guidance for the preparation and initial setup of the STC and supporting equipment. Section 10.2 provides the guidance for STC fuel loading. Section 10.3 provides the guidance for STC on-site transfer. Section 10.4 provides the guidance for STC fuel unload. Section 10.5 provides guidance for performing maintenance and responding to off-normal events. Equipment specific operating details such as valve manipulation and Transporter operation are not within the scope of this report and will be prepared based on the specific equipment selected by the users and the configuration of the site.

The steps contained herein describe acceptable methods for performing STC loading, unloading, and transfer operations. These procedures may be altered to allow alternate methods and operations to be performed in parallel or out of sequence as long as the general intent of the guidance is met. Users may select alternate configurations, equipment and methodology may be selected to accommodate their specific needs provided that the intent of this guidance is met. The steps provided in this chapter, equipment-specific operating instructions, and plant working procedures will be utilitized to develop the site specific written, loading and unloading procedures.

Technical and Safety Basis for Loading and Unloading Procedures The procedures herein are developed for the loading, unloading, and transfer of spent fuel in the STC. The activities involved in loading of spent fuel in the STC, if not carefully performed, may present risks. The design of the STC, including these steps, the ancillary equipment and the plant Technical Specifications, serve to minimize risks and mitigate consequences of potential events. To summarize, consideration is given in the loading, unloading and transfer systems and procedures to the potential events listed in Table 10.0.1. The primary objective is to reduce the risk of occurrence and/or to mitigate the consequences of the event. The steps contain Notes, Warnings, and Cautions to notify the operators to upcoming situations and provide additional information as needed. The Notes, Warnings and Cautions are purposely bolded and boxed and immediately precede the applicable steps. In the event of an extreme abnormal condition (e.g.,

cask drop) the user shall have appropriate procedural guidance to respond to the situation. As a minimum, the procedures shall address establishment of emergency action levels; implementation of emergency action program; establishment of personnel exclusions zones; monitoring of radiological conditions; actions to mitigate or prevent the release of radioactive materials; and recovery and planning, execution, and reporting to the appropriate regulatory agencies, as required.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289, 10-1 Rev. I

Table 10.0.1 OPERATIONAL CONSIDERATIONS POTENTIAL EVENTS METHODS USED TO ADDRESS EVENT Cask Drop During Cask lifting and handling equipment is designed to ANSI N14.6.

Handling Operations Procedural guidance is given for cask handling, inspection of lifting equipment, and proper engagement to the trunnions.

Cask Tip-Over during The STC and the HI-TRAC while supported on horizontal loading and unloading surfaces has been demonstrated by analysis to be stable and not tip-over from postulated events.

Contamination spread Processing systems are equipped with exhausts that can be from cask process directed to the plant's processing systems or depressurization exhausts underwater.

Damage to fuel assembly Fuel assemblies always remain covered with water and are never cladding from oxidation subjected to air or oxygen during loading and unloading operations.

Ignition of combustible Ignition sources are not used for STC processing and gases can mixtures of gas (e.g., be controlled by ventilation.

hydrogen) during STC handling Excess dose from failed STC gas sampling allows operators to determine the integrity of fuel assemblies the fuel cladding; this allows preparation and planning for failed fuel. Failed fuel assemblies (i.e. assemblies that are not intact) and/or damaged fuel are not permitted for transfer in the STC.

Excess dose to operators The procedures provide ALARA Notes and Warnings when radiological conditions may change.

Excess generation of The STC / HI-TRAC system uses process systems that minimize radioactive waste the amount of radioactive waste generated. Such features include smooth surfaces for ease of decontamination efforts, prevention of avoidable contamination, and procedural guidance to reduce decontamination requirements.

Fuel assembly misloading Procedural guidance is given to perform assembly selection event verification and a post-loading visual verification of assembly identification prior to installation of the STC lid. A physical device (cell-blocker) and temperature monitoring during loading are used as preventative measures.

Load Drop Rigging and procedural guidance are provided for all lifts of heavy loads and meet the NUREG 0612, Section 5.1.6 requirements. The vertical cask drop during lifting with the VCT has been evaluated. The VCT has redundant drop protection during cask transport.

STC carrying hot particles Procedural guidance is given to radiologically survey the STC out of the SFP prior to removal from the SFP followed by wash down of the I STC when removing from the SFP HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-20942 89 10-2 Rev. I

10.1 STC Preparation and Setup 10.1.1 STC Inspections and Checkout

1. Perform general visual inspection of basket and canister for damage, degradation or foreign material that would prevent the fuel assembly from seating. Repair coatings as required per manufacturer's instructions.

2. Inspect seals and sealing surfaces to ensure that surface will effect the required seal and replace/repair as required per Holtec Operation and Maintenance manual.

3. Maintenance requirements for the seal are documented in Chapter 8, Section 8.5.

4. Inspect and lubricate bolting, replace as required per Table 8.5.1.

5. Ensure trunnions and special lifting devices have been inspected or load tested in accordance with ANSI N 14.6. Visually inspect trunnions and apply approved lubrication.

10.1.2 HI-TRAC Inspections and Checkout

1. General maintenance requirements for the HI-TRAC are documented in HI-STORM 100 FSAR, Chapter 9[K.A]. The HI-TRAC seals and sealing surface maintenance requirements are documented in Chapter 8, Section 8.5 of this report.

2. Perform general visual inspection of HI-TRAC and Solid Top Lid for damage or degradation. Repair coatings as required per manufacturer's instructions.

3. Inspect seals and sealing surfaces to ensure that surface will effect the require seal and replace/repair as required.

4. Inspect and lubricate bolting, replace as required per Table 8.5.1.

5. Ensure that trunnions have been inspected or load tested in accordance with ANSI N 14.6.

Visually inspect trunnions and apply approved lubrication.

6. Inspect HI-TRAC internal cavity for presence of foreign material and remove as required.

Note:

Inspection and installation of empty STC in the HI-TRAC may occur at any location or be performed at any time prior to use as long as the following steps are performed.

10.1.3 Preparation and Setup for Use

1. Place STC and HI-TRAC in the preparation area. Perform appropriate inspection as listed in section 10.1.1 and 10.1.2 above.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-3 Rev. I

2. Ensure that the instrumentation defined in Table 10.1.1 has current calibration and is available for use.

3. If necessary, remove the HI-TRAC Top Lid by removing the top lid bolts and using the lift sling that meet the requirements of NUREG 0612, Section 5.1.5. Store Top Lid and bolts in a site-approved location.

ALARA Warning:

Replacement of the Pool Lid may only be performed when the HI-TRAC is empty.

4. If necessary, install the HI-TRAC Pool Lid with seal and tighten bolting.

5. If necessary, remove the HI-TRAC Pool Lid drain connection hardware. Install a threaded pipe plug with approved seal compound in the HI-TRAC Pool Lid drain.

6. Attach the Lift Lock or Lift Cleats to the STC Lid as needed. Attach the Lift Yoke to the STC Lid and engage them with the STC trunnions.

7. Using an overhead crane, engage the crane hook with the STC lifting device.

8. Place the empty STC inside the HI-TRAC.

9. Fill the HI-TRAC annulus between the STC with demineralized water to a maximum elevation of 3 inches below the top of the STC.

10. Check for water leakage at the HI-TRAC Pool Lid seal and the plugged drain connection.

Replace seal or clean sealing surfaces as required.

11. Remove the STC Lid bolting.

12. Disengage Lift Yoke arms from the STC trunnions and using the overhead crane. Lift and remove the STC Lid.

13. Using a suitable pumping system, fill the STC with SFP water to an elevation of 3 inches from the top of the STC.

14. Ensure that the STC seal is seated properly on the STC sealing surface.

15. Ensure that the STC Lid sealing surface is clean and free of debris.

16. Open the Lift Yoke arms and install the STC Lid aligned to allow the Lift Yoke arms to engage the STC trunnions.

17. Ensure the Lift Yoke arms are engaged with the STC trunnions.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-4 Rev. 1

18. Slowly lift the STC Lid without lifting the STC, verify engagement of the Lift Yoke arms with the STC trunnions.

19. Install the STC Lid bolting and hand tighten.

Table 10.1.1 STC INSTRUMENTATION

SUMMARY

FOR LOADING AND UNLOADING OPERATIONS Instrument Function Contamination Survey Monitors fixed and non-fixed contamination levels.

Instruments Dose Rate Monitors / Survey Monitors dose rate and contamination levels and ensures Equipment proper function of shielding. Ensures assembly debris is not inadvertently removed from the spent fuel pool during STC removal.

Pressure Gauges Ensures correct pressure during leak testing.

Temperature Indicating Monitors water temperature above a fuel assembly in the Device loaded STC prior to removal from the SFP.

Torque Wrench Ensures proper bolting preload for the STC and HI-TRAC lid bolting.

10.2 STC Fuel Loading 10.2.1 Placement of STC in the SFP

1. Ensure that the SFP water temperature is < 160 OF and the SFP cooling system is in operation.

2. Ensure that the SFP boron concentration is sufficiently above the TS minimum requirement of 1000 ppm to allowfor dilution from addition of demineralized water for wetting down of STC.

3. Engage the Lift Lock with the crane hook by engaging the connection pin through the hook center hole.

4. Slowly lift the STC from the HI-TRAC one or two inches.

5. Ensure that the Lift Yoke arms are engaged with the STC trunnions and the STC Lid bolting is hand tight.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-5 Rev. I

6. Using the overhead crane, lift the empty STC and position over the SFP cask loading area.

7. Wet down the STC and handling equipment with demineralized water.

8. Lower the STC into the SFP cask handling area.

9. Prior to completely submerging the STC, perform the following:

a. If necessary, open the vent valve and using a suitable pumping system, fill the STC with SFP water through the drain connection until water exits the vent connection.

b. Remove the STC Lid bolting and open the vent and drain connections.

10. Continue lowering the STC to the SFP floor. Ensure STC is a minimum of 8 inches from existing fuel racks.

11. Ensure no load exists on the crane hook.

12. Disengage the Lift Yoke arms from the STC trunnions. Using an underwater viewing device, verify that the Lift Yoke arms have been disengaged from the trunnions.

13. Slowly raise the crane and STC Lid to the SFP surface.

14. Wash down the STC Lid and lifting equipment with demineralized water for contamination control and store in a designated location.

10.2.2 Fuel Handling and Loading into STC Notes:

Fuel assembly misloadings are defined as:

Loading any Type 1 fuel assembly (per TS Figure 3.7.18-1) under Configuration 1; or

Loading any fuel assembly in the 4 center cells under Configuration 2.

Loading a fuel assembly recently irradiated fuel assembly.

The operational controls are based on the following:

" Fresh fuel assemblies can be distinguished visually from spent fuel assemblies, and

Except for fresh assemblies for a reactor core reload, there are only a small number of fuel assemblies in the pool that correspond to Type 1.

Use of cell blocking device.

" Temperature monitoring of the STC cells.

1. For each fuel transfer, the following steps will be performed.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-6 Rev. I

a. The fuel assemblies intended for transfer will be characterized to ensure compliance with IP3 Technical Specification 3.7.18 and IP2 Technical Specification 3.7.13. This characterization will be performed in accordance with approved Reactor Engineering procedures.

b. Fuel move sheets will be developed using the results of the fuel characterization process. The fuel move sheets will be independently checked by a qualified Reactor Engineers as required by Reactor Engineering procedures and become the approved load plan.

c. Prior to removal of a fuel assembly from its SFP storage rack location, fuel handling personnel will verify and peer check that the fuel assembly physical ID number correctly corresponds to the fuel assembly ID number specified in the approved load plan.

d. Peer checking will be performed by fuel handling personnel after placement in the STC to ensure the cell location is as designated in the approved load plan.

e. At the completion of loading, a video recording of the fully loaded STC will be performed and peer checked to ensure that the STC has been properly loaded in accordance with the approved load plan.

f. Prior to transfer of the loaded STC to IP2, an independent check of the video recording will be performed by a qualified Reactor Engineer to demonstrate that the STC has been properly loaded in accordance with the approved load plan.

2. For loading in Configuration 1:

a. Place a small cell blocker device on top of the center of the basket that prevents assemblies being loaded in the four center cells of the STC basket.

b. Load fuel in the eight outer cell locations in accordance with fuel move sheets.

c. Independently verify by visual inspection that none of the loaded assemblies in the STC are a fresh fuel assembly.

d. Verify that all assemblies in the pool that would not qualify as Type 2 assemblies are still present and in their expected location in the SFP.

e. Remove the cell blocker device

f. For each of the four locations in the center of the basket:

(l) Load the fuel assembly in accordance with fuel move sheets.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-7 Rev. I

(2) Independently verify by visual inspection that none of the loaded assemblies in these STC cells are a fresh fuel assembly.

(3) Verify that all assemblies in the pool that would not qualify as Type 2 assemblies are still present and in their expected location.

3. For loading in Configuration 2:

a. Place a small cell blocker device on top of the center of the basket that prevents assemblies being loaded in the four center cells of the STC basket.

b. Load fuel in the eight outer cell locations in accordance with fuel move procedures.

c. Remove the cell blocker device.

4. Using a remote temperature measuring device, take and record a water temperature reading at the top of each STC basket cell.

5. Compare temperature readings. Large differences in temperature between STC cells may indicate a misloading of a recently irradiated fuel assembly. Recheck the fuel selection sheets versus the fuel assembly loaded and take appropriate steps, if necessary; to ensure all fuel is loaded in accordance with the technical specifications.

6. Independently verify by visual inspection that the correct fuel assemblies have been loaded. An underwater viewing device may be used to verify and record the fuel identification numbers.

10.2.3 Removal of STC from SFP and placement in HI-TRAC

1. Ensure that the SFP boron concentration is sufficiently above the TS minimum requirement of 1000 ppm to allow addition of demineralized water for wetting down of STC.

2. Ensure that the Lift Lock and Lift Yoke are installed on the STC Lid.

3. Connect the overhead crane hook to the Lift Lock by engaging the connection pin through the hook center hole.

4. Move the STC Lid over the SFP and align with the STC.

5. Wet down the STC Lid and lifting equipment with demineralized water.

6. Ensure that the Lift Yoke arms are opened.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-8 Rev. 1

Note:

An underwater viewing device may be used for monitoring underwater operations.

7. Using an underwater viewing device, ensure that the STC seal is in place and the sealing surface is free of debris.

8. Using the alignment pins and Lift Yoke aligned with the STC trunnions, lower the STC Lid onto the STC.

9. Engage the Lift Yoke arms with the STC trunnions.

ALARA Note:

Activated debris may have settled on the STC during fuel loading. The top surface should be kept under water until a preliminary dose rate survey clears the STC for removal. Users are responsible for any water dilution considerations.

10. Slowly raise the STC to just below the SFP surface.

11. Perform a radiological survey of the top area of the STC to check for hot particles and remove as required.

12. Visually verify that the STC Lid is properly seated. If not, lower the STC, reinstall the Lid and repeat as necessary.

13. Raise the STC to allow access to the lid bolting and the vent and drain connections.

14. Wash down the STC and lifting equipment with demineralized water for contamination control.

15. With the STC Lid just above the SFP water surface, install the STC Lid bolting hand tight.

16. Connect a suitable pumping system to the STC drain connection and remove a small amount of water from the STC to avoid spilling water during handling.

17. Continue raising the STC while washing it down with demineralized water

18. Perform radiological surveys and compare to the expected dose rates as referenced in Chapter 7.

19. Place the STC into the HI-TRAC.

20. Disconnect the crane from the Lift Lock. Remove the Lift Lock from the STC Lid and store in a designated area.

21. Tighten STC Lid bolting to the specified preload.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-9 Rev. I

ALARA Warnin:

Water flowing from the STC may carry activated particles and fuel particles. Apply appropriate ALARA practices around the drain line.

22. Using a suitable pumping system, connect to the STC drain and pump out water until suction is lost to create the required air space.

23. Pressurize the STC to 55 +51-0 psig with air or nitrogen and hold for 10 minutes.

Perform a leak test of the STC Lid seal per ANSI N14.5, Section A.5.7 Soap Bubble test method. The acceptance criterion is no observed bubbles caused by leakage.

24. Depressurize the STC and remove the pressure test equipment.

25. Close the STC vent and drain connections.

26. Fill, as necessary, the STC/HI-TRAC annulus space with demineralized water to a maximum elevation of 3 inches below the top of the STC flange.

27. Ensure that the HI-TRAC seal is in place and the sealing surfaces are free of debris.

28. Place the Solid Top Lid on the HI-TRAC and install the bolting, tighten bolting to the specified preload.

29. Pressurize the HI-TRAC to 30 +51-0 psig with air or nitrogen and hold for 10 minutes.

Perform a leak test of the HI-TRAC Solid Top Lid seal per ANSI N14.5, Section A.5.7 Soap Bubble test method. The acceptance criterion is no observed bubbles caused by leakage.

30. Perform a leak check of the HI-TRAC Pool Lid and Drain plug. The acceptance criterion is no observed leakage.

31. Depressurize the HI-TRAC and remove the pressure test equipment.

32. Close the HI-TRAC vent connection.

33. Perform surface dose rate measurements for the HI-TRAC. Compare the measured dose rates with calculated dose rates to ensure they are less than the expected dose rates in Chapter 7.

10.3 HI-TRAC / STC Movement Note:

The haul path conditions must satisfy the requirements documented in the engineering change package for roadway evaluations and upgrades. The haul path will be evaluated and upgraded to meet the static and dynamic load condition for use of the VCT movement of the HI-TRAC with the STC loaded with fuel assemblies.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-10 Rev 1

10.3.1 Haul Path Inspection and Controls

1. Perform a inspection of the haul path to ensure that the conditions for fuel transfer are met.

2. Verify that transitory combustibles and hazards are not within haul path exclusion area.

3. Ensure administrative controls have been initiated to control transitory combustibles and hazards.

4. Ensure deliveries and vehicle traffic have been suspended in the haul path exclusion area.

5. Ensure no maintenance activities that involve the use of ignition sources (welding, burning, or grinding) or involve the use of flammable or combustible liquids are being performed in the haul path area and buildings within the exclusion area.

6. Verify the National Weather Service does not predict severe weather during the expected transfer period and average ambient temperatures.

7. Ensure a hot work qualified fire watch is assigned to the VCT/loaded cask and has an inspected 20 lb. type ABC fire extinguisher.

8. Ensure radiological controls are established in accordance with plant procedures and program requirements.

9. Ensure plant security controls are established in accordance with the security plan and implementation procedures.

10. Ensure that plant operations / shift manager notifications have been made.

10.3.2 Movement of loaded HI-TRAC on LPT

1. Air Pallets/Pads/Bearings at Unit 3

a. Ensure transport area is clean and free of debris

b. Operate the air pads per the manufacturer's instructions.

c. Use suitable prime mover connected to the HI-TRAC to control the load and move along the designated haul path from the FSB to the VCT lift point.

d. Secure the air pad system and remove the prime mover.

2. Hilman Roller Cart at Unit 2

a. Ensure haul path and channels are clean and free of debris.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-11 Rev. 1

b. Ensure the Hilman Roller Cart maintenance and inspection has been performed per the manufacturer's instructions.

c. Operate the Hilman Roller Assembly to transport the loaded HI-TRAC from the VCT lift point to the FSB.

10.3.3 Movement of loaded HI-TRAC with VCT

1. Ensure that VCT maintenance and inspection, in accordance with the manufacturer's instructions, has been performed.

2. Ensure that the VCT lateral supports for the HI-TRAC are installed.

3. Ensure that the HI-TRAC Lift Links have been inspected or load tested per ANSI N14.6 requirements.

4. Ensure that haul path requirements and controls in section 10.3.1 have been established.

5. Align the VCT over the HI-TRAC and engage the Lift Links with the HI-TRAC trunnions.

Note:

The loaded HI-TRAC maximum lift height limit is 6 inches. Appropriate surface support material may be used to lift the HI-TRAC in increments of less than 6 inches. When the VCT redundant drop protection is engaged, the maximum lift height limit does not apply.

6. Without exceeding the 6 inch maximum lift height limit, lift the HI-TRAC off the LPT with the VCT to the transport height and engage the locking pins.

a. Use a measuring device to ensure HI-TRAC does not exceed the 6 inch lift height limit.

b. Install suitable support surface material between the horizontal surface and the HI-TRAC.

c. Repeat as necessary to lift the HI-TRAC to the required height.

7. Install the VCT cask support strap around the HI-TRAC and engage the hydraulic tensioner.

8. Following the established haul path and controls, move the loaded HI-TRAC from the lift point at IP-3 to the setdown point at IP-2.

9. Remove the VCT cask support strap.

10. Without exceeding the maximum lift height limit, remove the locking pins and lower the loaded HI-TRAC onto the LPT.

11. Disengage the Lift Links from the HI-TRAC trunnions.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-12 Rev. 1

12. Move the VCT to the designated storage location.

10.4 STC Fuel Unloading 10.4.1 Placement of loaded STC in SFP

1. Ensure that the SFP boron concentration is sufficiently above the TS minimum requirement of 2000 ppm to allow for dilution from the STC water and addition of demineralized water for wet down of STC.

2. Ensure that the HI-TRAC has been positioned in the FSB where the gantry crane canister hoist can access the STC.

3. Depressurize the HI-TRAC by connecting a suitable hose to the HI-TRAC Solid Top Lid vent, routing it to the SFP or plant approved location, and opening the vent.

4. Remove the HI-TRAC Solid Top Lid bolting, remove the Lid, and store in a designated area.

Caution:

Oxidation of neutron absorber panels contained in the STC may create hydrogen gas while the STC is filled with water. Additionally, radiolysis of the water may occur in high flux conditions creating additional combustible gases. Appropriate monitoring for combustible gas concentrations shall be performed prior to, and during STC depressurization. The space below the STC lid may be purged with inert gas.

5. Depressurize the STC by connecting a suitable hose to the STC vent, routing it to the SFP or plant approved location, and opening the vent.

6. Loosen the STC Lid bolting and allow the Lift Yoke arms to engage the STC trunnions.

7. Install the Lift Cleats and Lift Cleat Adapter on the STC Lid.

8. Ensure theLift Yoke arms are engaged with the STC trunnions.

9. Connect the gantry crane canister hoist to the STC through the Lift Cleat Adapter.

10. Slowly lift the STC from the HI-TRAC one or two inches.

11. Verify that the Lift Yoke arms have engaged the STC trunnions and ensure the STC Lid bolting is hand tight.

12. Continue to lift the STC and place over the SFP cask handling area.

13. Wet down the STC and handling equipment with demineralized water.

14. Lower the STC into the SFP until the STC Lid is just above the SFP water surface.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-13 Rev. I

15. Prior to completely submerging the STC, perform the following:

a. If necessary, open the vent valve and using a suitable pumping system, fill the STC with SFP water through the drain connection until water exits the vent connection.

b. Remove the STC Lid bolting and open the vent and drain connections.

16. Continue lowering the STC and place on the SFP floor in the cask handling area. Ensure STC is a minimum of 8 inches from existing fuel racks.

17. Ensure no load exists on the crane hook.

18. Disengage the Lift Yoke arms from the STC trunnions. Using an underwater viewing device, verify the Lift Yoke arms have been disengaged from the trunnions.

19. Slowly raise the crane and STC Lid to the SFP surface.

20. Wash down the STC Lid and lifting equipment with demineralized water for contamination control and store in a designated location.

10.4.2 Unloading of STC

1. Ensure that fuel selection has been performed and is in compliance with the STC TS 3.7.15 and the fuel move sheets have been approved in accordance with plant procedures.

2. Using the approved fuel move sheets, move each fuel assembly from the STC and place it in the designated SFP rack cell.

3. Perform independent visual verification of each fuel assembly location.

4. If a fuel assembly cannot be placed in the designated SFP rack cell, return the fuel assembly to its former STC cell location per the fuel move sheets. Contact Reactor Engineering to obtain new fuel move procedure.

10.4.3 Removal of STC from SFP and placement in HI-TRAC

1. Ensure that the SFP boron concentration is sufficiently above the TS 3.7.12 minimum requirement of 2000 ppm to allow for dilution from addition of demineralized water for wash down of STC.

2. Verify no fuel assemblies are present in the STC.

3. Ensure that the Lift Cleats, Lift Cleat Adapter, and Lift Yoke are installed on the STC Lid.

4. Connect the gantry crane canister hoist to the Lift Cleat Adapter.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-14 Rev. 1

5. Move the STC Lid over the SFP and align with the STC.

6. Wet down the STC Lid and lifting equipment with demineralized water.

7. Ensure that the Lift Yoke arms are opened.

Note:

An underwater viewing device may be used for monitoring underwater operations.

8. Using the alignment pins and Lift Yoke aligned with the STC trunnions, lower the STC Lid onto the STC.

ALARA Note:

Activated debris may have settled on the STC during fuel unloading. The top surface should be kept under water until a preliminary dose rate scan clears the STC for removal. Users are responsible for any water dilution considerations.

9. Engage the Lift,Yoke arms with the STC trunnions.

10. Slowly raise the STC to just below the SFP surface. Survey the top area of the STC to check for hot particles and remove as required.

11. Visually verify that the STC Lid is properly seated. If not, lower the STC, reinstall the Lid and repeat as necessary.

12. Continue to raise the STC to allow access to the Lid bolting and the vent and drain connects.

13. Wash down the STC and lifting equipment with demineralized water for contamination control.

14. With the STC Lid just above the SFP water surface, install the STC Lid bolting hand tight.

15. Connect a suitable pumping system to the STC drain connection and remove a small amount of water from the STC to avoid spilling water during handling.

16. Continue raising the STC while washing down it down with demineralized water.

17. Perform radiological surveys and decontamination as required.

18. Place the STC into the HI-TRAC.

19. Disconnect the crane canister hoist from the Lift Cleat Adapter.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-15 Rev. 1

20. Remove the Lift Cleat Adapter and Lift Cleats from the STC Lid and store in the plant designated area.

21. Ensure the STC Lid bolting is installed and tightened to the preload requirements.

22. Ensure that the STC vent and drain connects are closed.

23. ' Install HI-TRAC Solid Top Lid and tighten the bolting to the preload requirements.

10.5 Maintenance and Off-Normal Events 10.5.1 Crane Operational Event

1. In the event of a crane hang-up or loss of power, perform one the following:

a. Restore power to the crane.

b. Manually lower the load to a safe location which will ensure the STC is in an analyzed condition. This is either in the SFP or in the HI-TRAC. The main hoist lowering, bridge, and trolley movement can be manually performed using the crane manufacturer's maintenance and operations instructions.

10.5.2 STC Water Inventory Control

1. During preparation of the loaded STC for transfer, the water level shall be maintained at no more than 9 inches below the top of the STC.

2. To ensure water levels are maintained, a suitable SFP water source may be connected to the STC vent and water added until water exits the STC drain connection.

3. Verify STC water inventory daily.

4. -Once the STC has been sealed and tested, water inventory verification is NOT required.

10.5.3 VCT Breakdown

1. Maintain transfer requirements and controls per section 10.3.1 until the VCT is repaired.

2. Using the VCT manufacturer's instructions, perform the required maintenance to restore VCT operations.

10.5.4 Vertical Cask Drop Recovery Plan

1. Inspect the HI-TRAC external surfaces for damage and the ability to handle the HI-TRAC with the VCT and/or the LPT. Make repairs as required for handling.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-16 Rev. I

2. Perform surveys and implement HI-TRAC unloading controls as defined in Section 10.5.5 for a potential failed fuel assembly.

3. Inspect the STC external surfaces for damage and the ability to handle the STC with the lifting devices. Make repairs as required for handling.

4. Perform surveys and implement STC unloading controls as defined in Section 10.5.5 for a potential failed fuel assembly.

10.5.5 Potential Damaged Fuel Assembly ALARA Note:

A gas sample analysis is performed to determine the condition of the fuel cladding in the STC.

The gas sample may indicate that fuel with damaged cladding is present in the STC. The results of the gas sample test may affect personnel protection and how the gas is processed during HI-TRAC and STC depressurization.

1. Connect radiological gas sampling equipment to the HI-TRAC vent. Route discharge from sampling equipment to plant processing system.

2. Based on result of gas sample analysis, establish the radiological controls needed for gas handling and radiation exposure controls.

3. Depressurize the HI-TRAC per Section 10.4.

4. Connect radiological gas sampling equipment to the STC vent. Route discharge from sampling equipment to plant processing system.

5. Based on result of gas sample analysis, establish the radiological controls needed for gas handling and radiation exposure controls.

6. Depressurize the STC per Section 10.4.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 10-17 Rev. I

CHAPTER 11: REFERENCES The following generic industry and Holtec produced references may have been consulted in the preparation of this document. Where specifically cited, the identifier is listed in the text or table. Active Holtec Calculation Packages which are the repository of all relevant licensing and design basis calculations are annotated as "latest revision". Submittal of the latest revision of such Calculation Packages to the USNRC and other regulatory authorities during the course of regulatory reviews is managed by the company's Configuration Control system.

A. United States Code of Federal Regulations

[A.A] U.S. Code of Federal Regulations, Title 10 "Energy", Chapter I "Nuclear Regulatory Commission", Part 50 "Domestic Licensing of Production and Utilization Facilities", January 2006.

[A.B] U.S. Code of Federal Regulations, Title 10 "Energy", Chapter I "Nuclear Regulatory Commission", Part 20 "Standards for Protection Against Radiation", January 2006.

[A.C] United States Code of Federal Regulations Title 10 "Energy", Chapter I "Nuclear Regulatory Commission", Part 72 "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste", January 2006.

B. American National Standards Institute (ANSI) Documents

[B.A] ANSI N45.2.1 - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants.

[B.B] ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Handling of Items or Nuclear Power Plants (During the Construction Phase).

[B.C] ANSI N45.2.6 - Qualifications of Inspection, Examination, and Testing Guide 1.58).

Personnel for Nuclear Power Plants (Regulatory

[B.D] ANSI N45.2.8, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Plants.

[B.E] ANSI N45.2. 11, Quality Assurance Requirements for the Design of Nuclear Power Plants.

[B.F] ANSI N45.2.12, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-1 Rev. 1

[B.G] ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Equipment Materials and Services for Nuclear Power Plants (Regulatory Guide 1.123).

[B.H] ANSI N45.2.15 Hoisting, Rigging, and Transporting of Items For Nuclear Power Plants.

[B.I] ANSI N45.2.23 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (Regulatory Guide 1.146).

[B.J] ANSI N16.9-75 Validation of Calculation Methods for Nuclear Criticality Safety.

[B.K] ANSI/ANS 8.1 (N16.1) - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors

[B.L] ANSI/ANS 8.17, Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors

[B.M] ANSI N45.2 - Quality Assurance Program Requirements for Nuclear Facilities - 1971

[B.N] ANSI N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974

[B.O] ANSI N45.2.10 - Quality Assurance Terms and Definitions -1973

[B.P] ANSI/ANS 57.2 - Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants - 1983.

[B.Q] ANSI N14.6 - American National Standard for Special LiftingDevices for Shipping Containers Weighing 10,000 pounds (4500 kg) or more for Nuclear Materials- 1992

[B.R] ANSI/ASME N626-3, Qualification and Duties of Personnel Engaged in ASME Boiler and Pressure Vessel Code Section III, Div. 1, Certifying Activities

[B.S] ANSI N14.6-1993, "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials", June 1993.

[B.T] ANSI N14.5-1997, "American National Standard for Radioactive Materials

- Leakage Tests on Packages for Shipment", June 1997.

[B.U] ANSI/ANS-6.1.1-1977, "American National Standard Neutron and Gamma-Ray Flux-to-Dose Rate Factors", June 1977.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-2 Rev. 1

[B.V] ANSI/ANS 57.9-1992, "Design Criteria for an Independent Spent Fuel Storage Installation (Dry Type)", Re-affirmed 2000.

C. USNRC Standard Review Plans (NUREG)

[C.A] NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants",

USNRC, Washington D.C., 1980

[C.B] NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants", Section 3.5.1.4, Rev. 2, July 1981

[C.C] NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants", Section 9.1.2, Rev. 3, July 1981

[C.D] NUREG-1536, "Standard Review Plan for Dry Cask Storage Systems",

USNRC, Washington D.C., January 1977

[C.E] NUREG-1567, "Standard Review Plan for Spent Fuel Dry Storage Facilities", USNRC, Washington D.C., March 2000

[C.F] NUREG-1617, "Standard Review Plan for Transportation Packages for Spent Nuclear Fuel", USNRC, Washington D.C., 2000

[C.G] NUREG/CR-0497, "A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior", Revision 2, USNRC, Washington D.C., August 1981

[C.H] NUREG/CR-5661, "Recommendations for Preparing the Criticality Safety Evaluation for Transportation Packages", USNRC, Washington D.C., April 1997

[C.I] NUREG/CR-6322, "Buckling Analysis of Spent Fuel Basket", USNRC, Washington D.C., May 1995

[C.J] NUREG/CR-6407, "Classification of Transportation Packaging and Dry Spent Fuel Storage System Component According to Important to Safety",

USNRC, Washington D.C., February 1996

[C.K] NUREG/CR-6760, "Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit", USNRC, Washington D.C., 2002

[C.L] NUREG/CR-6800, "Assessment of Reactivity Margin and Loading Curves for PWR Burnup Credit Cask Designs", USNRC, Washington D.C., 2003

[C.M] NUREG/CR-68 11, "Strategies for Application of Isotopic Uncertainties in Burnup Credit", USNRC, Washington D.C., 2003 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-3 Rev. 1

[C.N] NUREG/CR-1864, "A Pilot Probabilistic Risk Assessment of a Dry Cask Storage System at a Nuclear Power Plant", USNRC, Washington D.C.,

2007

[C.O] USNRC, NUREG-1437, Supplement 38, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Regarding Indian Point Nuclear Generating Unit Numbers 2 and 3, Draft Report, December 2008

[C.P] Anderson, B.L. et al. Containment Analysis for Type B Packages Used to Transport Various Contents. NUREG/CR-6487, UCRL-ID-124822.

Lawrence Livermore National Laboratory, November 1996.

D. USNRC Regulatory Guides

[D.A] Regulatory Guide 1.59, "Design Basis Floods for Nuclear Power Plants",

Revision 1, April 1976

[D.B] Regulatory Guide 3.61, "Standard Format for a Topical Safety Analysis Report for a Spent Fuel Storage Cask", USNRC, Washington D.C.,

February 1989

[D.C] Regulatory Guide 7.9, "Standard Format and Content of Part 71 Applications for Approval of Packaging for Radioactive Material",

Revision 2, USNRC, Washington D.C., March 2005

[D.D] Regulatory Guide 7.10, "Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material", Revision 2, USNRC, Washington D.C., March 2005

[D.E] Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposure at Nuclear Power Stations will be As Low As Reasonably Achievable", USNRC, Washington D.C., June 1978

[D.F] Regulatory Guide 8.10, "Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Reasonably Achievable",

Revision l-R, USNRC, Washington D.C., May 1997

[D.G] RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 2 Proposed)

[D.H] RG 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility of Boiling and Pressurized Water Reactors

[D.I] RG 1.28 - (ANSI N45.2) - Quality Assurance Program Requirements

[D.J] RG 1.29 - Seismic Design Classification (Rev. 3)

[D.K] RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Material HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-4 Rev. 1

[D.L] RG 1.38 - (ANSI N45.2.2) Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants

[D.M] RG 1.44 - Control of the Use of Sensitized Stainless Steel

[D.N] RG 1.58 - (ANSI N45.2.6) Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel

[D.O] RG 1.61 - Damping Values for Seismic Design of Nuclear Power Plants, Rev. 0, 1973

[D.P] RG 1.64 - (ANSI N45.2.1 1) Quality Assurance Requirements for the Design of Nuclear Power Plants

[D.Q] RG 1.71 - Welder Qualifications for Areas of Limited Accessibility

[D.R] RG 1.74 - (ANSI N45.2.10) Quality Assurance Terms and Definitions

[D.S] RG 1.85 - Materials Code Case Acceptability - ASME Section 3, Div. 1

[D.T] RG 1.88 - (ANSI N45.2.9) Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records

[D.U] RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis

[D.V] RG 1.122 - Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components

[D.W] RG 1.123 - (ANSI N45.2.13) Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants

[D.X] RG 1.124 - Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, Revision 1, 1978

[D.Y] RG 3.4 - Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities

[D.Z] RG 3.41 - Validation of Calculational Methods for Nuclear Criticality Safety, Revision 1, 1977

[D.AA] RG 8.8 - Information Relative to Ensuring that Occupational Radiation Exposure at Nuclear Power Plants will be as Low as Reasonably Achievable (ALARA)

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-5 Rev. 1

[D.BB] DG-8006, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants"

[D.EE] U.S. Nuclear Regulatory Commission, "Atmospheric Dispersement Models for Potential Accident Consequence Assessments at Nuclear Power Plants,"

Regulatory Guide 1.145, February 1989.

[D.CC] IE Information Notice 83 Fuel Binding Caused by Fuel Rack Deformation

[D.DD] RG 8.38 - Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June, 1993 E. Interim Staff Guidance (ISG) Documents

[E.A] SFST-ISG-1, "Damaged Fuel"

[E.B] SFST-ISG-2, "Fuel Retrievability"

[E.C] SFST-ISG-3, "Post Accident Recovery and Compliance with 10 CFR 72.122(1)"

[E.D] SFST-ISG-4, Revision 1, "Cask Closure Weld Inspections"

[E.E] SFST-ISG-5, Revision 1, "Confinement Evaluation"

[E.F] SFST-ISG-6, "Establishing Minimum Initial Enrichment for the Bounding Design Basis Fuel Assembly(s)"

[E.G] SFST-ISG-7, "Potential Generic Issue Concerning Cask Heat Transfer in a Transportation Accident"

[E.H] SFST-ISG-8, Revision 2, "Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks"

[E.I] SFST-ISG-9, Revision 1, "Storage of Components Associated with Fuel Assemblies"

[E.J] SFST-ISG-10, Revision 1, "Alternatives to the ASME Code"

[E.K] SFST-ISG-1 1, Revision 3, "Cladding Considerations for the Transportation and Storage of Spent Fuel"

[E.L] SFST-ISG-12, Revision 1, "Buckling of Irradiated Fuel Under Bottom End Drop Conditions"

[E.M] SFST-ISG-13, "Real Individual" HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-6 Rev. 1

[E.N] SFST-ISG- 14, "Supplemental Shielding"

[E.O] SFST-ISG- 15, "Materials Evaluation"

[E.P] SFST-ISG- 16, "Emergency Planning"

[E.Q] SFST-ISG-17, "Interim Storage of Greater Than Class C Waste"

[E.R] SFST-ISG-18, "The Design/Qualification of Final Closure Welds on Austenitic Stainless Steel Canisters as Confinement Boundary for Spent Fuel Storage and Containment Boundary for Spent Fuel Transportation"

[E.S] SFST-ISG-19, "Moderator Exclusion Under Hypothetical Accident Conditions and Demonstrating Subcriticality of Spent Fuel Under the Requirements of 10 CFR 71.55(e)"

[E.T] SFST-ISG-20, "Transportation Package Design Changes Authorized Under 10 CFR Part 71 Without Prior NRC Approval"

[E.U] SFST-ISG-21, "Use of Computational Modeling Software"

[E.V] SFST-ISG-22, "Potential Rod Splitting Due to Exposure to an Oxidizing Atmosphere During Short-Term Cask Loading Operations in LWR or Other Uranium Oxide Based Fuel"

[E.W] Interim Staff Guidance-5, Revision 1, "Normal, Off-Normal and Hypothetical Dose Estimate Calculations", June 18, 1999 F. Other USNRC Documents

[F.A] USNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978, and Addendum dated January 18, 1979.

[F.B] ANSI N210-1976, "Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations" (contains guidelines for fuel rack design).

[F.C] USNRC Bulletin 96-04: "Chemical, Galvanic or Other Reactions in Spent Fuel Storage and Transportation Casks", July 5, 1996

[F.D] USNRC Information Notice 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly Sealed Basket", May 1996

[F.E] USNRC ASLB, "Final Partial Initial Decision on F-16 Aircraft Accident Consequences", Docket No. 72-22-ISFSI, ASLB# 97-732-02-ISFSI, dated 2/24/2005.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-7 Rev. 1

[F.F] Certificate of Compliance for Spent Fuel Storage Casks, Certificate 1014, Docket 72-1014, Amendment No. 5.

[F.G] NRC Letter (Boska) to Entergy (IPEC) dated June 11, 2009 (ML091520167)

G. American Society of Mechanical Engineers (ASME) Codes

[G.A] ASME Boiler and Pressure Vessel Code,Section II, Parts A - Ferrous Material Specifications, 2007.

[G.B] ASME Boiler and Pressure Vessel Code,Section II, Parts D - Properties, 2007.

[G.C] ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NB - Class 1 Components, 2007.

[G.D] ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NF - Supports, 2007.

[G.E] ASME Boiler and Pressure Vessel Code,Section III, Appendices, 2007

[G.F] ASME Boiler and Pressure Vessel Code,Section V - Nondestructive Examination, 2007

[G.G] ASME Boiler and Pressure Vessel Code,Section IX - Welding and Brazing Qualifications, 2007.

[G.H] ASME Boiler and Pressure Vessel Code,Section XI - Rules for Inservice Inspection of Nuclear Power Plant Components, 2007.

[G.I] ASME Steam Tables, 3 rd Edition (1977)

[G.J] ASME NQA-2-1989, Quality Assurance Requirements for Nuclear Facility Applications.

[G.K] ASME Boiler and Pressure Vessel Code,Section IX - Welding and Brazing Qualifications, latest Edition.

[G.L] ASME Boiler and Pressure Vessel Code NCA3550 Requirements for Design Documents, latest Edition.

[G.M] ASME Boiler and Pressure Vessel Code NCA4000 - Quality Assurance, latest Edition.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-8 Rev. 1

[G.N] ASME NQA-1, Requirements for the establishment and execution of quality assurance programs for the siting, design, construction, operation, and decommissioning of nuclear facilities, 1994 Edition.

H. Other Standards

[H.A] American Society for Testing and Materials, ASTM A 352-93, "Ferritic and Martensitic Steel Castings for Pressure-Containing Parts Suitable for Low-Temperature Service"

[H.B] American Society for Nondestructive Testing, ."Personnel Qualification and Certification in Nondestructive Testing", Recommended Practice No. SNT-TC-1A, December 1992

[H.C] American Society for Testing and Materials, ASTM E 1003-05, "Standard Test Method for Hydrostatic Leak Testing"

1. Metamic Reports

[I.A] EPRI Report 1003137, "Qualification of Metamic for Spent Fuel Storage Applications", Palo Alto, CA, October 2001.

[I.B] Holtec Report No. HI-2043215, Latest Revision, "Sourcebook for Metamic Performance Assessment".

[I.C] USNRC Letter, Alexion (NRC) to Anderson (ANO), "Arkansas Nuclear One, Units 1 And 2 - Review Of Holtec Report Regarding Use of Metamic in Fuel Pool Applications", June 17, 2003

[I.D] California Consolidated Technology Inc., "Metamic 6061 +40% Boron Carbide Metal Matrix Composite Test", August 2001.

J. Industry Reports

[J.A] A. Luksic, "Spent Fuel Assembly Hardware: Characterization and 10CFR61 Classification for Waste Disposal", PNL-6909 Vol. 1, Pacific Northwest Laboratory, June 1989

[J.B] EPRI, Greer et al., NP-5128, PNL-6054, UC-85, "The TN-24P Spent Fuel Storage Cask: Testing and Analyses", April 1987

[J.C] Nuclear Systems Materials Handbook, Volume 1 Design Data, TID 26666, Vol. 1, Oak Ridge National Laboratory

[J.D] Peckner and Bernstein, "Handbook of Stainless Steels," First Ed., 1977 (pp 16-17).

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-9 Rev. 1

[J.E] Craig and Anderson, "Handbook of Corrosion Data," ASM International, First Ed., 1995.

K. Holtec Licensing Reports

[K.A] HI-2002444, Latest Revision, "Final Safety Analysis Report for the HI-STORM 100 Cask System", USNRC Docket 72-1014

[K.B] HI-2012610, Latest Revision, "Final Safety Analysis Report for the Holtec International Storage, Transport, And Repository Cask System (HI-STAR 100 Cask System), USNRC Docket 72-1008

[K.C] HI-951251, Latest Revision, "Storage, Transport, and Repository Cask System (HI-STAR Cask System) Safety Analysis Report", USNRC Docket 71-9261

[K.D] HI-2089327, Latest Revision, "Licensing Report for'Increased Storage Capacity for Indian Point Unit 2 Spent Fuel".

[K.E] HI-2073710, "Safety Analysis Report on the HI-STAR 60 Transport Package", USNRC Docket 71-9336, Latest Revision.

[K.F] HI-89327, Revision 2, "Licensing Report for Reracking Indian Point Unit 2 Spent Fuel Pool".

L. Holtec Technical Reports

[L.A] HI-2043215, Proprietary, S. Turner PhD, "Sourcebook for Metamic Performance Assessment", 2004

[L.B] Holtec International Quality Assurance Program, Latest Approved Revision.

[L.C] Quality Assurance Documentation Package for ANSYS (Version 5.3 and Higher), HI-2012627, Revision 3, 2005.

[L.D] Quality Assurance Documentation Package for LS-DYNA 3D, HI-961519, Revision 5, 2006.

[L.E] Holtec International Final Safety Analysis Report for HI-STORM 100 Cask System, Revision 7, dated August 9, 2008.

[L.F] Holtec International Report HI-2084179, "Criticality Safety Evaluation of the IP-3 Shielded Transfer Canister", Latest Revision

[L.G] Holtec International Report HI-2084109, "Shielding Design Calculations of Shielded Transfer Canister for Indian Point 3", Latest Revision HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-10 Rev. 1

[L.H] Holtec International Report HI-2084146, "Thermal Hydraulic Analysis of IP3 Shielded Transfer Canister", Latest Revision

[L.I] Holtec International Report HI-2084118, "Shielded Transfer Canister Structural Calculation Package", Latest Revision M. Computer Codes (Public Domain)

[M.A] ANSYS, Version 5.7, 7.0, 9.0, or 11.0

[M.B] CASMO-4 (A), Version 1.13.04 (Unix), 2.05.03 (Windows), or 2.05.14 (Windows)

[M.C] CASMO-3 (A), Version 4.4 or 4.7

[M.D] LS-DYNA3D (A), Version 936, 940, 950, 960, 970, or 971

[M.E] FLUENT (A) Version 4.32, 4.56, 5.5, 6.1.18, or 6.2.16 or 6.3.26.

[M.F] KENO-5A (A), Version 4.3 or 4.4

[M.G] MCNP (A), Version 4A, 4B, or 5

[M.H] ORIGENS (Scale), Version 4.3 or 4.4

[M.I] SAS2H (Scale), Version 4.3 or 4.4 N. Handbooks/Texts

[N.A] Crane Manufacturer's Association of America (CMAA), Specification #70, 1988

[N.B] Shah, M.J., Klymyshyn N.A., and Kreppel B.J., "HI-STAR 100 Spent Fuel Transport Cask Analytic Evaluation for Drop Events", Packaging, Transport, and Security of Radioactive Materials, Vol. 18, No. 1, W.S.

Maney & Sons (2007).

[N.C] "Mechanical Design of Heat Exchangers and Pressure Vessel Components", by K.P. Singh and A. I. Soler, Arcturus Publishers, Cherry Hill, New Jersey, 1100 pages, hardbound (1984).

0. Holtec Patents

[O.A] "Fuel Basket", Patent No. 5,898,747, April 27, 1999.

[O.B] "HI-STORM Overpack", Patent No. 6,064,710, May 16, 2000.

[O.C] "Duct Photon Attenuator", Patent No. 6,519,307B1, February 11, 2003.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-11 Rev. 1

[O.D] "HI-TRAC Operation", Patent No. 6,587,536B1, July 1, 2003.

[O.E] "Cask Mating Device (Hermetically Sealable Transfer Cask)", Patent No.

6,625,246B 1, September 23, 2003.

[O.F] "Improved Ventilator Overpack", Patent No. 6,718,000B2, April 6, 2004.

[O.G] "Below Grade Transfer Facility", Patent No. 6,793,450B2, September 21, 2004.

[O.H] "HERMIT (Seismic Cask Stabilization Device)", Patent No. 6,848,223B2, February 1, 2005.

[0.1] "Cask Mating Device (Divisional)", Patent No. 6,853,697, February 8, 2005.

[O.J] "Davit Crane", Patent No. 6,957,942B2, October 25, 2005.

[O.K] "HI-STORM IOOU", Patent No. 7,068,748B2, June 27, 2006.

[O.L] "Forced Helium Dehydrator", Patent No. 7,096,600B2, August 29, 2006.

[O.M] "Below Grade Transfer Facility", Patent No. 7,139,358B2, November 21, 2006.

[O.N] "Forced Gas Flow Canister Dehydration", Patent No. 7,210,247B2, May 1, 2007.

P. Standard Review Plan

[P.A] SRP 3.2.1 - Seismic Classification

[P.B] SRP 3.2.2 - System Quality Group Classification

[P.C] SRP 3.7.1 - Seismic Design Parameters

[P.D] SRP 3.7.2 - Seismic System Analysis

[P.E] SRP 3.7.3 - Seismic Subsystem Analysis

[P.F] SRP 3.8.4 - Other Seismic Category I Structures (including Appendix D),

Technical Position on Spent Fuel Rack

[P.G] SRP 3.8.5 - Foundations for Seismic Category I Structures, Revision 1, 1981

[P.H] SRP 9.1.2 - Spent Fuel Storage, Revision 3, 1981 HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-12 Rev. I

[P.I] SRP 9.1.3 - Spent Fuel Pool Cooling and Cleanup System

[P.J] SRP 9.1.4 - Light Load Handling System

[P.K] SRP 9.1.5 - Heavy Load Handling System

[P.L] SRP 15.7.4 - Radiological Consequences of Fuel Handling Accidents Q. AWS Standards

[Q.A] AWS D 1.1 - Structural Welding Code, Steel

[Q.B] AWS D1.3 - Structure Welding Code - Sheet Steel

[Q.C] AWS D9.1 - Welding of Sheet Metal

[Q.D] AWS A2.4 - Standard Symbols for Welding, Brazing and Nondestructive Examination

[Q.E] AWS A3.0 - Standard Welding Terms and Definitions

[Q.F] AWS A5.12 - Tungsten Arc-Welding Electrodes

[Q.G] AWS QC1 - Standards and Guide for Qualification and Certification of Welding Inspectors R. Thermal-Hydraulic References

[R.A] Baumeister, T., Avallone, E.A. and Baumeister III, T., "Marks' Standard Handbook for Mechanical Engineers," 8th Edition, McGraw Hill Book Company, (1978).

[R.B] Rohsenow, W.M. and Hartnett, J.P., "Handbook of Heat Transfer,"

McGraw Hill Book Company, New York, (1973).

[R.C] Creer et al., "The TN-24P Spent Fuel Storage Cask: Testing and Analyses,"

EPRI NP-5128, PNL-6054, UC-85, (April 1987).

[R.D] Rust, J.H., "Nuclear Power Plant Engineering," Haralson Publishing Company, (1979).

[R.E] Kern, D.Q., "Process Heat Transfer," McGraw Hill Kogakusha, (1950).

[R.F] "A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior," NUREG/CR-0497, (August 1981).

[R.G] "Spent Nuclear Fuel Effective Thermal Conductivity Report," US DOE Report BBA000000-01717-5705-00010 REV 0, (July 11, 1996).

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-13 Rev. 1

[R.H] ASME Boiler and Pressure Vessel Code,Section II, Part D, (1995).

[R.I] Jakob, M. and Hawkins, G.A., "Elements of Heat Transfer," John Wiley &

Sons, New York, (1957).

[R.J] ASME Steam Tables, 3rd Edition (1977).

[R.K] "Nuclear Systems Materials Handbook, Vol. 1, Design Data", ORNL TID 26666.

[R.L] "Scoping Design Analyses for Optimized Shipping Casks Containing 1-,

2-, 3-, 5-, 7-, or 10-Year-Old PWR Spent Fuel", ORNL/CSD/TM-149 TTC-0316, (1983).

[R.M] "Qualification of METAMIC for Spent-Fuel Storage Application", EPRI Report 1003137, (October 2001), EPRI, Palo Alto, CA.

[R.N] "Sourcebook for METAMIC Performance Assessment", Holtec Report HI-2043215, Holtec International, Marlton, NJ, 08053.

[R.O] USNRC Docket no 72-1027, TN-68 FSAR & Docket no 72-1021 TN-32 FSAR.

[R.P] Hagrman, Reymann and Mason, "MATPRO-Version 11 (Revision 2) A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior," NUREG/CR-0497, Tree 1280, Rev. 2, EG&G Idaho, August 1981.

[R.Q] "Effective Thermal Conductivity and Edge Conductance Model for a Spent-Fuel Assembly," R. D. Manteufel & N. E. Todreas, Nuclear Technology, 105, 421- 440, (March 1994).

[R.R] "Thermal Measurements in a Series of Large Pool Fires", Gregory, J.J1 et.

al., SAND85-1096, Sandia National Laboratories, (August 1987).

[R.S] "Pressure Loss Characteristics for In-Cell Flow of Helium in PWR and BWR MPC Storage Cells", Holtec Report HI-2043285, Rev. 5.

S. Shielding References

[S.A] Forrest B. Brown, Russell D. Mosteller, and Avneet Sood, "Verification of MCNP5," Proceedings of M&C 2003: A Century in Review, A Century Anew, Gatlinburg, Tennessee (April 2003).

T. Economic and Environmental Considerations References HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-14 Rev. 1

[T.A] USNRC Generic Letter 78-11, OT Position for Review and Acceptance of Spent Fuel Storage and Handling Application

[T.B] USAEC, Final Environmental Statement related to operation of Indian Point Nuclear Generation Plant Unit Number 2, Docket 50-247, September 1972, Volume I and II

[T.C] USNRC, Final Environmental Statement related to operation of Indian Point Nuclear Generation Plant Unit Number 3, Docket 50-286, February 1975, Volume I and II

[T.D] Electric Power Research Institute, Report No. NP-3380, Cost Comparison for On-Site Spent Fuel Storage Options, May 1984.

[T.E] Indian Point Energy Center, Indian Point IP-2, Updated Final Safety Analysis Report, Revision 20

[T.F] Indian Point Energy Center, Indian Point IP-3, Updated Final Safety Analysis Report, Revision 2

[T.G] Indian Point Energy Center, Indian Point IP-2, Improved Technical Specifications

[T.H] Indian Point Energy Center, Indian Point IP-3, Improved Technical Specifications

[T.I] Entergy Nuclear Management Manual, Procedure EN-RP-l 10, ALARA Program

[T.J] U.S. EPA, Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, DE89-011065, 1988.

[T.K] Northeast Technology Corporation Report NET- 173-02, "Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis."

[T.L] Northeast Technology Corporation Report NET-173-01, "Criticality Analysis for Soluble Boron and Bumup Credit in the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks."

U. Structural References

[U.A] Timoshenko and Gere, Theory of Elastic Stability, McGraw Hill, 1961.

[U.B] IPEC HI-STORM 100 Cask System 72.212 Evaluation Report.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-15 Rev. 1

[U.C] Holtec Report HI-2084118, Revision 1, "STC Structural Calculation Package".

[U.D] Holtec Report HI-2094345, Revision 0, "Analysis of a Postulated HI-TRAC IOOD Drop Accident During Spent Fuel Wet Transfer Operation".

[U.E] Rabinowicz, E., "Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company, 1976.

V. Criticality References

[V.A] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

[V.B] J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).

[V.C] "Lumped Fission Product and Pml48m Cross Sections for MCNP," Holtec Report HI-2033031, Rev 0, September 2003 (proprietary).

[V.D] M. Edenius, K. Ekberg, B.H. Forss6n, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1, Studsvik of America, Inc. and Studsvik Core Analysis AB (proprietary).

[V.E] D. Knott, "CASMO-4 Benchmark Against Critical Experiments", SOA-94/13, Studsvik of America, Inc., (proprietary).

[V.F] D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/12, Studsvik of America, Inc., (proprietary).

[V.G] L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T. Collins, August 19, 1998.

[V.H] ANS-8.1/N16.1-1975, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," April 14, 1975.

[V.1] "Recommendations for Addressing Axial Bumup in PWR Burnup Credit Analyses", ORNL/TM-2001/273, NUREG/CR-6801, USNRC Office of Nuclear Regulatory Research, March 2003.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL REPORT HI-2094289 11-16 Rev. 1

ENCLOSURE 3 TO NL-09-076 AFFIDAVIT EXECUTED PURSUANT TO 10 CFR 2.390 GOVERNING THE PROPRIETARY INFORMATION INCLUDED IN THE HOLTEC LICENSING REPORT Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

mEum. Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797-0900 HOLTEC INTERNATIONAL Fax (856) 797-0909 July 2, 2009 Mr. Robert W. Walpole Licensing Manager Indian Point Energy Center 450 Broadway GSB Second Floor Licensing Buchanan, NY 10511-0249 Document ID: 1775012

Subject:

Information to Support Licensing Submittal on Inter-Unit Fuel Transfer

Dear Mr. Walpole:

Holtec is pleased to approve the release of the following information to the United States Nuclear Regulatory Commission (USNRC): : Standard CD Labeled "Attachment 1 to Holtec Letter 1775012" containing one PDF file: HI-2094289R1 .pdf (Proprietary) : Standard CD Labeled "Attachment 2 to Holtec Letter 1775012" containing one PDF file: HI-2094289R1 -nonprop.pdf (Non-Proprietary)

We require that you include this letter along with the attached affidavit pursuant to 10CFR2.390 when submitting Attachment I to the USNRC.

Please do not hesitate to contact me at 856-797-0900 x 687 if you have any questions.

Sincerely, Tammy Morin Licensing Manager Holtec International Page lof I

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 1775012 Non-Proprietary Attachment AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Tammy S. Morin, being duly sworn, depose and state as follows:

(1) I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2) The information sought to be withheld is Holtec report HI-2094289R1 contained in Attachment 1 to Holtec letter Document ID 1775012, containing Holtec Proprietary information.

(3) In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(1) for "trade secrets and commercial or financial information obtained from 'a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information", and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC.Cir. 1992),

and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir.

1983).

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U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 1775012 Non-Proprietary Attachment AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a, 4.b, and 4.e above.

(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have 2 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 1775012 Non-Proprietary Attachment AFFIDAVIT PURSUANT TO 10 CFR 2.390 been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.

Access to such documents within Holtec International is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec's competitor to copy our technology and offer it for sale in competition with our company, causing us financial injury.

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U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 1775012 Non-Proprietary Attachment AFFIDAVIT PURSUANT TO 10 CFR 2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

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U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 1775012 Non-Proprietary Attachment AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEYý )

) ss.

COUNTY OF BURLINGTON)

Tammy S. Morin, being duly sworn, deposes and says:

That she has read the foregoing affidavit and the matters stated therein are true and correct to the best of her knowledge, information, and belief.

Executed at Marlton, New Jersey, this I st day of July, 2009.

Tammy S. Morin Holtec International Subscribed and sworn before me this day of ,2009.

NOGTARY P.BLC OF NEW JE.R6L 01 TA ofmRYMARIA C MA6-AprO 25,2 ssofn*lxpires 5 of 5

v t e Letter Sequence Request
TAC:L24299, Control Room Habitability (Open)
Initiation Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting Questions 16 November 2010 28 September 2011 9 November 2011 Answers 23 August 2011 28 October 2011 11 January 2012 5 October 2010 28 October 2010 28 July 2011 28 October 2011 15 December 2011 2 March 2012 Results Approval... Other: ML092090289, ML103080113, ML11243A220, ML120740550
MONTHYEAR2011-11-09 [Table View]

ML091940178

Text

SUMMARY

Title:

SUMMARY

SUMMARY

7.0 INTRODUCTION

SUMMARY

Subject:

Dear Mr. Walpole:

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